Gas-Cooled Reactors and Their Applications

IAEA-TECDOC-436

GAS-COOLED REACTORS AND THEIR APPLICATIONS

PROCEEDINGS OF A TECHNICAL COMMITTEE MEETING GAS-COOLEN O D REACTOR THEID SAN R APPLICATIONS ORGANIZED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY AND HELD IN JÜLICH, 20-23 OCTOBER 1986

A TECHNICAL DOCUMENT ISSUED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1987 GAS-COOLED REACTOR THEID SAN R APPLICATIONS IAEA, VIENNA, 1987 IAEA-TECDOC-436

Printe IAEe th Austri An i y d b a October 1987 FOREWORD

Technological development r gas-coolefo s d reactor present a e ar st bein C(>2-cooler gfo larga carrieK n U o e e t scaldth ou d Advancen i e d Gas-Cooled Reactors Higr (AGRsfo h d Temperatur)an e Gas-Cooled Reactors (HTGRs) in the Federal Republic of Germany, United States of America, Japan and USSR.

l programmeAl directee sar d towards constructio plantsf o n e Th . mose th t r advancefa y b s i d programmK efforU e th t n witi e h several commercial Hagnox Reactor and Advanced Gas-Cooled Reactor units in operatio d furthean n re constructio th AGR n i s r commissionino n g phase. HTGRs are in operation in the USA and in the Federal Republic of Germany. Othe R specifirHT c wor beins ki g carrieAustrian i t ou d , China and Switzerland. At the IAEA the International Working Group on Gas-Cooled Reactors (IWGGCR bees )ha n establishe 197n i d 8 in order to promote an exchange of information on gas-cooled reactor developmen stimulato t d tan e international cooperation framewore th n I . k f thio s working group, other countrie e followinar s developmene th g f to GCR technology in order to investigate its possible benefits for national energy supply e IWGGCTh .recommendes Rha Agence th conveno o t yt d e this Technical Committe Gas-Coolen eo d Reactor theid an s r Applications, which s attendewa mory b d e tha0 participant20 n s fro countrie5 2 m d an s International Organizations.

Agence Th gratefus i yGovernmene th Federao e lt th f to l Republif o c Germany and to the Nuclear Research Centre Jiilich for their hospitable and efficient arrangements. This volume contain paperl al s s presente meetinge th t a d . EDITORIAL NOTE

preparingIn this materialpress,the International the for staff of Atomic Energy Agency have mounted paginatedand originalthe manuscripts submittedas authorsthe givenby and some attention presentation.the to The views expressed papers,the statementsin the general the made and style adoptedare the responsibility of the named authors. The views do not necessarily reflect those of the govern- ments of the Member States or organizations under whose auspices the manuscripts were produced. The use in this book of particular designations of countries or territories does not imply any judgement publisher,the legalby the IAEA, to the status as suchof countries territories,or of their authorities and institutions or of the delimitation of their boundaries. The mention specificof companies theirof or products brandor names does implynot any endorsement or recommendation on the part of the IAEA. Authors are themselves responsible for obtaining the necessary permission to reproduce copyright material from other sources. Please be aware that all the Missing Pages thin i s document were originally blank pages CONTENTS

Introductio Summard nan y ...... 7 .

OVERVIE STATUE TH GAS-COOLEF F WSO O D REACTORD SAN THEIR PROSPECTS (Session A)

Magnox reactor advanced an s d gas-cooled reactors ...... 1 1 . A.F. Pexton High temperature gas-cooled reactors ...... 29 SchultenR.

EXPERIENCE WITH GAS-COOLED REACTORS (Sessio) nB

Development of a high output AGR ...... 43 B.A. Keen

Technical evolution and operation of French CO2 cooled reactors (UNGG) ...... 53 Y. Berthion Fort St. Vrain performance ...... 61 H.L. Brey AYR experience ...... 71 E. Ziermann THTR operation — The first year ...... 99 D. Schwarz

DESCRIPTION OF CURRENT GCR PLANT DESIGNS (Session C)

Restoratio CEGa f no B Magnox reacto fulo t r l power ...... l Il . R.J. Paynter, B.J. Roberts, P. T. Sawbridge The modular high-temperature gas-cooled reactor (MHTGR) ...... 123 A.J. Neylan Status of gas-cooled reactor development in the USSR ...... 135 V.N. Grebennik The nuclear electricity and heat generation using the VG-400 reactor ...... 141 V.V. Bulygin, V.E. Vorontsov, V.N. Grebennik, E.M. lonov, A.N. Protsenko, A.Ya. Stolyarevskij, S.V. Shishkin, Yu.K. Panov Current status of research and development of HTGR in Japan ...... ' 51 T. Hayashi Plant desig higa f nho temperature engineering test reacto JAERn i r I ...... 5 16 . O. Baba Status of the HTR-Module plant design ...... 179 I.A. Weisbrodt, W. Steinwarz, W. Klein Small nuclear powe gas-cooleR r plantGH M0 1 W s d: heating reactor ...... 9 19 . H. Schmitt HTR 100 industrial nuclear power plant for generation of heat and electricity ...... 213 Brandes,S. KohlW. HTR 500 — The basic design for commercial HTR power stations ...... 235 E. Baust, J. Schöning SAFETY ASPECTS (Session D)

Safety characteristic f curren o s...... R HT t 3 25 . W. Kroger The safety concept of the modular HTR ...... 267 G.H. Lohnen MHTGR licensing approach and plant response to off normal events ...... 283 A.J. Neylan, F.A. Silady Compariso f regulatorno y aspect differenn si t countries ...... 5 29 . HofinannK.

GAS-COOLED REACTOR APPLICATIONS (Sessio) nE

HTGR applications, prospectives and future development ...... 313 H. Barnert Perspective f HTGRo s chemican i s irod steed an nlan l industrie f Japao s n ...... 9 32 . K. Tsuruoka, T. Katamine, T. Miyasugi, R, Araki Israeli perspectiv HTGRn eo s ...... 3 34 . Barak,A. Beck,A. Greenspan,E. Szabo,J. Blumenau,L. Branover,H. A. El-Boher, E. Spero, S. Sukoriansky

GAS-COOLED REACTOR TECHNOLOGY (Session F)

Overvie MHTGS U f wo R base technology development program ...... 1 37 . J.E. Jones, Jr., P.R. Kasten Status and development of the German materials programme for the HTGR ...... 387 H. Nickel, F. Schubert, H. Schuster KV hige statud th K han f o stemperatur e component development ...... 7 40 . W. Jansing, H. Breitling, R. Candeli, H. Teubner HTR fuel development and qualification — Treatment of spent fuel ...... 429 G. Kaiser, HacksteinK. Characteristics of HTGR spherical fuel elements ...... 445 A.S. Chernikov, L.N. Permyakov, L.I. Mikhajlichenko, S.D. Kurbakov Self-sustaining thorium cycle in a high temperature graphite reactor ...... 461 K. Balakrishnan

USER'S PERSPECTIVES ON GAS-COOLED REACTORS (Session G)

Future applications of the high temperature reactor ...... 473 Klusmann,A. StelzerM. Financing models for HTR plants — Co-finaning, counter trade, joint ventures ...... 483 Bogen,J. StölzlD. Perspective HTGe th n Rso fro ...... A m US utilitie e th n si 9 49 . H.L. Brey, L.D. Mears Market prospect f HTRso i newlsh y industrialize developind dan g countries ...... 1 51 . Garriba,S. VivanteC.

Chairme Secretariad nan t ...... 1 52 . List of Participants ...... 523 INTRODUCTION AND SUMMARY

e invitatioth Governmene n O th Federae f o nth f to l Republif co German e Internationayth l Atomic Energy Agency convene Jiilichn i d , 20-23 October 1986 Technicae ,th l Committe Gas-Coolen eo d Reactord san their Applications. The meeting was hosted by the Nuclear Research Centre Jiilich and cosponsered by the Commission of the European Communities. The purpose of the meeting was to review and discuss the current status and recent progress made in the technology and design of gas-cooled reactor theid an s r applicatio electricitr nfo y generation, process steam and process heat production. The meeting was attended by more tha0 participantn20 s fro countrie5 m2 Internationad an s l Organizations presenting 34 papers. The technical part of the meeting was subdivided into 7 sessions: . A Overvie Statue th Gas-Coolef f wo so d Reactor d theian s r Prospects B. Experience with Gas-Cooled Reactors C. Description of Current GCR Plant Designs

D. Safety Aspects

E. Gas-Cooled Reactor Applications F. Gas-Cooled Reactor Technology G. User's Perspectives on Gas-Cooled Reactors meetine th f o rouna g d en d e tablth t eA discussio s organizenwa n i d order to summarize the meeting and to make recommendations for future activities. Gas-Cooled Reactors have been under development for more than three decades. Today there are more than 40 gas-cooled reactors in operation commissionine th n i r o g phas seven i e n Member IAEAe Stateth .f so Most of them are C02~Cooled Magnox Reactors and Advanced Gas-Cooled Reactors (AGRs) for electricity generation, having accumulated about 800 reactor years of operation. The Helium-Cooled High Temperature Gas-Cooled Reactors (HTGRs) are also for electricity generation, however, the long term developmen mosn i HTGe tts th i R goacountrie r fo l productioe th s n and application of process steam and process heat for chemical industries, coal conversion processes, nuclear steelmaking, etc. Currently there are also gas-cooled reactors of small sizes under design with reduced complexity, passive safety mechanisms and a high potential for modular design and shop fabrication. These reactors are not only appropriate for application in highly industrialized countries but also in countries with less developed industrial infrastructure. During the meeting the technology of gas-cooled reactors was presente discussed an d d together with their different applications. Particular emphasis was given to a review of their safety features, and the perspectives of users on this reactor line. After the meeting a tour too kTHTR-300e e pebbl placMW th d reactor o 0 be et e 30 e ,a th , n whici s i h commissioning phase and which has successfully reached the 100% power level.

The meeting concluded that gas-cooled reactors have reached a very hig hAGRse th leve particula n K i ,f maturityU lo e th e n I rth . Hinkley/Hunterston type design, have provee economib o t nd saf an ce electricity producers e successfuTh . l operatio f Higo n h Temperature Gas-Cooled Reactors in the USA and the Federal Republic of Germany and the commissionin THTR-30e th f go 0 have proven that this technologs i y availabl r electricitfo e y generatio d procesnan s steam productione Th . advantages of the system, i.e., high safety, high temperature potential and diverse applications, environmental acceptability and fuel saving have been confirme e operatioth y b d f prototypo n d demonstratioan e n plants. uniqua GCRe ar se too power d stealfo an r m production tooa , l whic margine bases th i h n f o safetdo s y inheren e desigth n i tn concept rather than on engineered safeguards, a feature which is of increasing interest in Agency Member States. Because of their specific properties, GCRe expectear s o havt d e very good international market chancen i s particula mediue th n i rm size d smalan d l power range. OVERVIE STATUE TH F WSO OF GAS-COOLED REACTORS AND THEIR PROSPECTS (Session A) MAGNOX REACTORS AND ADVANCED GAS-COOLED REACTORS

A.F. PEXTON South of Scotland Electricity Board, Glasgow, United Kingdom

Abstract

Commercial nuclea rU.Ke powebases th i . n CC>n i rdo 2 cooled graphite moderated reactors. This started around 1950 with the magnox programme having Calder HalChapelcrosd lan e th s sa forerunners. Nine twin-reactor commercial magnox stations were then built between 195 1970d 8firse an Th .t seven stations have steel pressure vessels with external boilers connectey db gas ducts e lashavo Th n integratettw ea . d design with core, boilers and gas circulators enclosed in pre-stressed concrete pressure vessels. The. magnox reactors have given good service and the older stations are now being subjected to careful review after 20 years operation.

Magnox was superseded by the AGR and five twin reactor stations to three separate and distinct designs, but all with integral gas circuits inside concrete pressure vessels, were orderen i d the period 196mos e 1970o Th t 5t .successfu thesef lo t a , Hinkley Point 'B' and Hunterston 'B' have now been in operation for ove years0 1 r . Their loa dpas e yearw factoth fe t r s rfo ordee bees th 80%f ro ha f no .

197n furtheo I tw 9 r twin reacto stationsR rAG , Heyshad an I mI Torness, were authorised. These wer eHinkleye baseth n do / Hunterston design with detailed improvements arising from operating experienc developmentd ean safetn i s y standards. Commissionin welw g no test le advancefirsar e s th td dreactoran s at each station are scheduled to come into operation in 1987. Wor s proceedekha doriginae closth o et l programm budgetd ean .

The Heysham II/Torness design now provides a standard base for on-going development of the AGR. A study is in hand which essentially replicates the main features, but by effective use marginf o s enables higher outpu achievede b o t .

The future depends on government policy following issue of the report of the inquiry into CEGB's application to build a PWR at Sizewell, as well as on the political environment following Chernobyl. Whatever the outcome, however, the AGR provides an attractive option with many inherently favourable safety characteristics. Whilst it is difficult to see an incentive for the U.K. to change to an alternative form of gas-cooled reactor technology, U.K. gas cooled reactor experience provides scop transfer fo e technologf ro HTGo yt R development Europn si e U.Se th . d an

11 e StructurTh Electricitf eo y Supply productio, UK e th d distribution I nan publif no c electricity supplie responsibilite th s i s State-ownef yo d monopolies. There are, for planning and commercial purposes, two integrated grid systems, one for England and Wales and one for Scotland but the interconnection capacity between them, some 1000MW, is adequate to allo power wfo r exchanges arising from economic schedulinf go e generatinth g plan botn to hbordere sideth f o s.

The Contribution from Nuclear Power

nucleae Th r stations servin overale gth l U.K. grid systee mar shown in Figure 1.

Dounrcay

Output Station Type (MWso) Berkeley Magnox 276 Bradwell Magnox 245 Calder Hall Magnox 200 Chapetcross Magnox 200 Dungeness A Magnox 424 Hinkley Point A Magnox 430 Hunterston A Magnox 300 Trawsfynydd Magnox 390 Sliewell Magnox 420 Ok) bury Magnox 434 Wytla Magnox 840 Hinkley Point B AGR 1040 Dungeness B AGR 3300 Harttepoo) AGR 3300 Heysham 1 AGR 3300 Heysham* II AGR 1230 Hunterston B AGR 1150 Hmkl« »S Poin A t Winfnth Tomess* AGR 1360 Oounreay PFR 250 Wlnfrrth SGHWR 100 •not )r*t commlMàorwd

FIG. 1. UK NUCLEAR POWER STATIONS.

Commercial nuclear powe bases i r n CC>2-cooledo d , graphite moderated reactors, starting witCaldee th h r Hall 'Magnox' type, which uses metallic natural uranium fuel with magnesium alloy cladding. Eight twin reactor commercial plants of this type, aldifferento t l , progressively large mord ran e developed designs, were e ElectricitbuilTh y tb y Supply Industrn i y England and Wales, and one station, Hunterston 'A' was built in Scotland, all between 1958 and 1970, providing a total of 4000MW(e). At one time there were five different industrial consortia competin desige th constructiod r nan fo g nucleaf no r power stations.

The commissioning of a 33MW(e) test bed for the Advanced Gas- Cooled Reactor (AGR) at Windscale in 1962 opened the way two years evaluation latea o t r competitivf no e tender AGRr fo sR ,PW d BWR an R system .adoptioo AG t Thi e d th .sle f no Fiv e stations, each with twin reactors of some 600MW(e), were ordered perioe ith n d 196 51970o t threo t , e separat d distincan e t designs from the three nuclear design and construction companies

12 still active in the UK. Two more twin reactor stations, CEGB's Heysham II and SSEB's Torness were authorised in 1979 based on CEGB's Hinkle ySSEB'd an Poin ' s'B tHuntersto n 'B' mose ,th t successfu three th ef l o origina l designs. Start-up of Heysham d TornesIan I schedules si 1987n i d .

At present nuclear power provides 19% of the total energy generated for public electricity in the UK but this will grow, with further plant now building and being commissioned, to more than 25% in the next few years. In Scotland the nuclear share is much higher presene ,th % increasint45 oveo withi% gt r60 n the next two years. These outputs are compared with that of other countries which hav significanea t nuclear contribution ni Figure 2. Actual information is shown for 1984 together with the OECD estimate r 1990sfo .

FIG NUCLEA. .2 R SHARE ELECTRICITF SO Y PRODUCTION.

The annual requirement for increased capacity in the UK is estimated at about 1200MW(e) per year (ie one plant per year) from the mid 1990s. In Scotland alone the corresponding requirement would be only one 1200MW(e) plant per decade but, in view of the long timescale required for planning and approvals, though alreads i t y being nex e giveth to nt plant .

Based on current cost estimates the advantages of AGR compared with new coal-fired plant for base load operation is at least 30%, expresse totan i d l cosunir tpe t sent out d coulan ,e db more depending on the extent of the provisions required for flue gas desulphurizatio productiow lo d nan nitrogef no n oxides. Clearly in the UK as a whole there is a strong economic incentiv increaso t e proportioe th e nucleaf plane no th tn i r mix. However fule th l f benefit,i nucleaa f so r ordering

13 achievedprogramme b o t e wilt ,i ear importane lb t that this twie restth n n foundationso reactoa f so r designe whicb n hca replicated without significant change constructioa d san n programme in which there is sufficient assurance of successive orders to allow proper planning of resources and career stability for trained and experienced personnel in the design, construction and manufacturing sectors.

The Magnox Programme

Calder Hall, in England, and Chapelcross, in Scotland, the forerunners of the magnox programme each comprise four reactors of 35 MW(e operatd ) an bar7 pressures t sga ea . Each reactos rha four separat circuitss ega , eachs containinga n ow s git circulato boiled ran r (Fig.3). Refuellin f carries gi of t dou load. The fuel element cans originally had simple circumferential fin thest sbu e were later modifiea o dt herringbone design which gave improved heat transfer. This, with other improvements, outpu enablet e increasene b e o t dth d to 50MW(e upratine reactor historicae )th pe th d d ran gan l load factor of Chapelcross are shown in figure 4. After 25 years of operation lifetime ,th e load statione factoth f ro , basen do rated output, is 85%.

Hunterston 'A' was the third twin reactor commercial magnox firss statioit td uni an ns commissioneU.K.e builtwa ,th n ti d in 1964. Each reacto eighs rha t separat circuits ega s with configuration show figurn carboe ni Th . e5 n dioxide pressurs i e 10 bars and the design net output is 150Mtf(e) per unit. Refuellin carries fult i ga lt loadou d usin refuellinga g machine having acces pressuro t s e vessel standpipes beloe wth reactor. The historical load factor of Hunterston 'A' is shown in figur . Aftee6 year2 r2 operatiof lifetimsa o s ha t eni load factor of 82% based on design output.

iZSâV///// ^//////. mm^

3 FIG. CROSS SECTION THROUGH CHAPELCROSS REACTOR BUILDING

14 200 _

MTtD ...r OUTPUT

100 ij970'7t ' n ' TJ ' 74 ' n ' n ' n ' nT n 'IMO' it ' R ' n ' n ' « ' X YEARS 100—-

rtcrai

N. u.

'iPOH l 1H 7l^ 1«7 « ' 6' ' 47J'74l7S ' «O * 5' ' >gBS ' O t >( lnl77l7i'7»l»lH'i»l«»li4l« YEARS A . ANNUA G LFI LOAD FACTOR CHAPELCROSF SO S BASE RATEN DO D OUTPUT

CHARGE MACHINE

GAS CIRCULATOR

5 HUNTERSTO G FI ^ MAGNO" N X

15 MWSO

DESIGN 300

OUTPUT

65/66 66/67 67/68 68/69 69/70 70/71 71/73 73/73 73/74 74/75 75/76 76/77 77/79 78/79 79/80 1980 81/82 83/83 83/84 84/85 85/86 /81 FINANCIAL YEARS 100 -T-

90 --

LOAD

FACTOR

ten.

i a / FINANCIAL YEABS 6 ANNUA. G L FI LOAD FACTOR HUNTERSTOF O S ' BASE 'A NDESIGN O D N OUTPUT

The first seven U.K." commercial magnox stations have steel reactor pressure vessels with external boilers connectey db ducts. A major change was introduced in CEGB's last two magnox stations, Oldbur Wylfad yan , which hav integraten ea d design circuits ga wit e whole th ,h th f eincludino g core, boilerd san gas circulators, enclosed in a prestressed concrete pressure vessel.

In circui196s ga 9a t corrosion proble encounteres mwa d which affecte magnol dal x stations majoe Th .r structural steel components had been manufactured from silicon killed mild steel with good corrosion resistance in moist CC>2 but certain other items had been made of more rapidly corroding rimming steel. e growt Th e face oxidf th ho corrodin f so n eo g washersr ,fo example somn i ed area,ha s buil locap tu l strains sufficieno t fail bolts. Deratin reactore th somf go f eo necessars swa y whils systematita c programm monitoringf eo , developmend tan modification was undertaken. At Hunterston 'A', for example, the mean duct outlet temperature was reduced from 390°C to 360°C givin statiot ne falga n li n output frodesige th m n valuf eo 300MW(e leso )t s than 280MW(e). Flow testscala n so e modef lo the reactor demonstrated that peak temperatures in the most critical areas could be reduced by leaving a few peripheral fuel channels emptyresula beew s outpue A no .n tth s restoretha o t d the original design value.

Magnox fuel is a success story. The design burn-up of commercial magnox fue initialls lwa y 3000MWd/Te, limitee th y db expected physical life of the fuel elements. The average

16 channel burnu beinw pno g achieve oves i d r 5,500MWD/Te. Over 2.5 million fuel elements hav beew eno n loaded inte th o commercial reactors with an average failure rate less than 0.1%, and the current failure rate is an order of magnitude less than this.

The Nuclear Installations Inspectorate decided that after 20 years operation that the magnox plants should be subject to a comprehensive safety review entire Th . e rang safetf eo s yha been studied including structural integrity of the pressure circuit, core and components, the ability of the reactor protection to respond to all types of reactor faults and the ability of structures to withstand external hazards. On Hunterston 'A' alone this review has absorbed over 100 man years of professional effort.

The AGR Programme

The commercial AGR programme adopted the concept of the integral design inside a concrete pressure vessel. stationR AlAG le th fiv sf eo ordere d between 196 d 1975an e 0ar operationaw no showind lan g encouraging potential. Hinkley Point 'B' and Hunterston 'B' have been generating for over ten years. Figure 7 compares cross sections of Hunterston 'B' and the subsequent design developmen Tornesst ta .

SHIELDING SA5 BAFFLE CONCRETE PRESSURE VESSEL

REACTOR BOILER DECAY HEAT BOILER GAS CIRCULATOR

SECONDARY SHUTDOWN ROOM

HUNTERSTO1 'B N TORNESS

FIG.7 COMPARISO REACTOF O N R PRESSURE VESSEL CROSS SECTIONS FOR HUNTERSTO D TOHNESAN ' 'B NS

17 The rated output of each unit at Hunterston was originally primarilW M limite0 46 uncertainty o yb dt corrosioe th o t s nya stee. Cr resistanc l% 9 boile e th r f e somo y tube b ed sminoan r deficiencies in fuel pin end-caps. With the progressive introductio detailef no d improvement authoriset ne e sth d rating has been raised to 575MW(e), close to the original design target. Further increases will be made as modified blades are fitte certai o dt turbine th f no e improves stagea d san d Stage2 reactor fuel elements are introduced.

The load factor has been progressively improved, first by eliminating sources of unreliability (mainly in the conventional plant) and, secondly, by establishing on-load refuelling as normal practice. It is necessary to reduce power to 30 per cent for refuelling at present to avoid the risk of damaging the graphite sleeves of the fuel assemblies but the Stage 2 fuel has mora e robust graphite sleev wild ean l allo refuelline wth g power to be raised. Further reductions of refuelling output losses will therefor achievee eb Stags da fuee2 l comee b o t s discharged.

Figure 8 shows progress in raising the output capability and load factor at Hunterston. Achievement at Hinkley has been similar.

Constructio earliee th f no r AGRs entailed fabricatio sitn no f eo such large components as the core supports, the gas baffles and the core restraint tanks. However Heyshar Tornessd ,fo an I mI , having established an agreed national design, it was decided to secure the benefits of factory manufacture and these components were transported fully assemble siteo manufacturinw dt Ne . g facilities were create produco dt onlt eno y these large fabrications but many others such as graphite blocks, gas circulators, boilers and fuelling equipment, all to very high

r-1200 ,____ _ 4100 10CU -1000 90_ _900 BO- BOO ' LOAD 70- -700 DECLARED FACTOR-- NETT % 60- .600 1 CAPABILITY 50_ -500 MW 40_ .400 30_ _300 1 i ————— ao_ -200 10- _100 0 1977/78 1978/79' 1979/80 19BO/B1 1981/88 198a/83 1983/84 1984/83 19B5/8É FINANCIAL YEAR

FIG.8 CAPABILITY AND LOAD FACTOR FOR HUNTERSTON B.

18 standard qualitf so witd yan h good productivity. Necessarily site-performed operations, suc thermalls ha y insulatine gth pressure vessel, were rationalised usin mose gth t up-to-date technique d full-scalsan e mock-up training facilities were provided for the operatives at site. Figure 9 shows the site construction sequence employed wit large hth e prefabricated components.

GAS BAFFLE

PRESSURE VESSEL LINER \

ATTACHED PENETRATIONS

PHASE 1 PHASE 2 LIFT-I BAFFLS NGA E ASSEMBLY FOLLOWEY B D ROLL-IN PRESSURE VESSEL LINER INSTALLATION OF PRESSURE VESSEL ROOF

LECTRICAD AN L INSTRUMENT •AREAS

PHASE 3 MAIN MECHANICAL/ELECTRICAL WORK FACES F16.9 A6R CONSTRUCTION PHASES AT TORNESS

Figure 10 shows the achievement in construction of the two units at Torness against the original programme set in 1979. The consent of the Nuclear Installations Inspectorate is expected loadine shortlth r fuef go yfo Reactounfuellee n li Th . r1 d commissioning tests have established the integrity of the pressure vessel and the gas circuit and other components of the plant have behave designes da d durin pressuriset gho flos dga w tests.

Hunterstoprototypa mane s th wa y d ' lessonean 'B n s learned from its constructio operatiod nan n have been incorporatee th n i d Torness project. Measurement actuaf so l operating performance at Hunterston have allowe ratee dth d outpu Tornesf to e b o st increasecenr pe t 7 withou y db t plant modifications. Once eth

19 AOVAHCf LATEST SYNCHRONISE START STAAT UNIT 1 OF HAIN OF KADI UNIT 1 CIVIL «ORKS civn. noms WJS 1330 FEE 1981 Vf 198?

STAUT OROBt Nil flou. LIFT IN COMPLETE COMPLETE COMPLETE COMBINED FULL DESIGN SAS CONSENT IH US flOttJEH CAS BOILER ENGINES! ING LOAD WfflK BAFFLE LUCH BAFFLE LOADING BAFFLE HYDRAULIC TESTS UNIT 1 KATERIAL TESTS TESTS OCT 197B JUN 1981, MAY 1982 KAR 1983 FE8 1984 M. 1985 AUS laps APR 1986 NAY 1987

PflE START CONSTRUCTION UYCTG SAFETY flEPOBT GRAPHITE TO «I uAN I960 FEB 1984 pLAhMB er? z-ç-o/ZTZ-Z-Z.: r r ACHIEVED ' ZLZ^C^i SYNCWONISE UNIT 2

MAR 1988 ROLL LIFT IN COMPLETE COMPLETE COMPLETE CONBtHB) FULL tN CAS BOILER GAS BOILER ENBPBïlDIG LOAD- LMER BAFFLE LOADING BAFFLE HYDRAULIC TESTS UNIT a UNIT 2 TESTS TESTS (AY 1983 MAR 1984 FE8 1985 JUL 1986 UK 1986 APR 1987 HAY 1988 LAYWSTARTB GRAPHITE FEB 1985 r1

1967 1988

FIG. 10. TORNESS CONSTRUCTION PROGRAM SHOWING COMPARISON OF ORIGINAL PLAN AND ACHIEVED DATES plant has settled down, trimming of the operating conditions could permit some further small increase. The expected outturn uni taccordancn i cos s i t e witoriginae th h l estimate.

The Torness specification incorporates a number of new requirements. Provision is made for improved access to component structured san s withi pressure nth e vesser lfo inspection and maintenance. Extra provision has also been made remotr fo e inspection. Figur 1 illustratee1 s these facilities.

INSID BAFFLS EGA E INSIDE VAULT

REPAIR AREA COMPONENTS REPAIR AREA COMPONENTS ISI ISI FEASIBLE FEASIBLE REMOTE OUTSIDE SHIELD AND REMOTE INSIDF EO GUIDE. VISUAL. PERIPHERAL OF PERIPHERAL VISUAL/ YES S BAFFLSA E TUBES. (AN ACCESS GUIDE MAN TO TUBES ACCESS CENTRE DOME PERIPHERAL fHffii ONLY HANGERS CENTRE TUBES REMOTE TAILPIPES VISUAL/ INSIDF EO GAS BAFFLE REMOTE REHEATER HEADERS MAN YES KNUCKLE. VISUAL/ CASING BAFFLE YES - SHIELD ACCESS DOME UPPER MAN GAS SEAL ANNULUS fi NEUTRON ACCESS TAILPIPES REMOTE KNUCKLE SHIELD MAIN PENETRATIONS VISUAL/ REMOTE BOILER TRANS JOINTS MAN YES INSIDF EO RESTRAINT VISUAL. CASINGS ACCESS GAS BAFFLE TANK AND LIMITED NO ' BOILER CYLINDER BAFFLE MAN ACCESS ANNULUS CASING SHIELD REMOTE C WALEN P L TO T A ABOVE PENETRATIONS VISUAL/ MAIN SEAL MAN YES BOILER RESTRAINTS ACCESS GRAPHITE REMOTE GAS SEAL CORE NO „ BOILER MAIN SEAL REMOTE CHANNELS VISUAL ANNULUS DIVISION VISUAL, AT BOILER PLATES MAN YES SUPPORTS SUPPORTS ACCESS DIAGRID. REMOTE TAILPIPES DIAGHID WEBS VISUAL/ YES - CIRCULATOR REMOTE SKIRT MAN SUB BOILER YES CONNECTION. ACCESS ANNULUS gaîus W PIPESD SS . SKIRT ACCESS

FIG. 11 AGR GAS CIRCUIT - IN SERVICE INSPECTION AND REPAIR

20 Experienc Hunterstot ea n 'B', afte year0 r1 operationf so , justifie expectatioe sth n that radiation levels withie nth vessel will permi entrn substantiaa tma n yo l scale throughout the Station's life. Figure 12 shows how the radiation dose acquired by Hunterston 'B' AGR workers compares with typical values in PWR stations.

Cracks appearin reactoe welde th th somn f gi so f r eo roof insulatio Hunterstot na afte' yearsn 'B n te r ' operation have been reporte Technicae th n i d l Press e thermaTh . l insulatiof no the pressure vessel around each fuel standpipe consists of

1.3-,

(PWRA US . 4 LOOP PLANT) 1.0 - r

MAN REM/ M.W. (E) Y L_ I I L_J UJ R SWEDEoucncPN Nn GERMAN (RINGHAL) S2

SIZEWELL 'B' L _. DESIGN TARGET

HUNTERSTON 'B' AGR

73 7< 1 4 9 7 5 8 7 76 7 7 81 82 63 84 YEARS FIG 12 COMPARISON OF OCCUPATIONAL DOSE PER UNIT OF ENERGY GENERATED FOR DIFFERENT REACTOR TYPES

mineral fibre blankets interleaved with stainless steel foils, the whole being retaine y stainlesdb s steel tube platesd san . Figure 13 shows the arrangements at Hunterston and the improved design at Torness. At Hunterston, certain welds, labelled A Figurn i hav, B 14 ed e an suffered crackin y thermagb l fatigue. A small number of these have been removed for inspection and testing. Research covering structural analysis and flow modellin w showinno s i gg that these welds have been subjecteo t d cyclic temperature variations due to instability of the local gas flow e solutio.Th modifient fi derive o t ds i d thermal plugs

21 TORNESS

SECONDARY RETENTION FOR MAIN COVER PLATE

L STANOPH>UE £ INSULATION «FUEL AMCMM. 9HOWMT YNO )

HUNTERSTON

SECONDARY RETENTIOB NFO MAIN COVtR PLATE DIRECTIOF O N 9 G vieFI N wI FUEL STAWDPIPE INSULATION (FUEL ASSEMBLY NOT SHOWN)

FIG13 ARRANGEMEN F INSULATIOO T N LINENO R ROOF

removable tth o e fuel assemblies. Non thi f eo s safet sha y connotations shore least th ,a tn i t term , because th f eo redundancy built intinsulatiodesige e th oth f no n restraint structure, and normal operations have continued with the consent Nucleae ofth r Installations Inspector.

Safety

Design safety guidelines have been adopted by the UK Electricity Supply Industry which include requirements to meet numerical radiological release targets e principaTh . m lai targeo t s i t

22 FUEL STANDPIPE

CONCRETE STANDPIPE INSULATI8H- COOLING INNER TUBE WATER PIPES

CERAMIC FIBRE INSULATION nuuROOrF INSULATION

MAIN COVER PLATE WELD WELD

WELD (ft——— WELD (£) BOTTOM CLOSURE RING SEALING RING WELD

FIG . 14 CROSS SECTION THROUGH BOTTOM OF FUEL STANDPIPE.

fo desiga r n which reduce totae sth l probabilitn a f yo uncontrolled releas activitf eo environmene th o t y leso t s than 10"^ per reactor year.

The design of the Torness safety systems resulting from the application of the design safety guidelines is extremely robust, with diverse provision maie th nn si function faul- s t sensing, reactor shutdown and post trip cooling - to maximise reliability and afford protection against common mode failures. There are mai diversd nan e guardlines maie Th n. guardlin solia s i ed state LADDIC system operating on two out of four logic while the diverse guardline is a two out of three system using relays. primare Th y shutdown system uses control rods inserted inte oth reactor core under gravity while, to meet the requirement for diversit safetl al n yi y functions secondar,a y shutdown system is provided which injects nitroge s intcore nga oth e from beneat reactore hth manuallA . y initiated reactor holddown system involving the insertion of boron beads from beneath the core is also incorporated.

Both guardlines will initiate the primary and secondary shutdown systems. However, only one system is required for reactor

23 shutdown and therefore operation of the secondary shutdown system is automatically inhibited once adequate control rods are inserted intcoree oth .

The post trip cooling systems have a high degree of diversity redundancyd an . They operate automaticall brino yt reactoe gth r into a shutdown mode. The plant is designed so that operator actio nrequiree b wil t lno d withi minute0 n3 faula f t so t bu the operator can, of course, act to support the operation if a safety system or post trip cooling does not operate correctly. hige th ho t therma e Du l inerti core moderatd th an e f ao e fuel rating the reactor is very tolerant to failure of the post trip cooling systems. This is illustrated by two examples.

First, los electricaf so l supplie scirculatos leadga o t s r speed run-down which in turn rapidly reduces gas flow to the reactor core. The reactor trips from gas circulator under-voltage or underspeed signals. The diesel generators start from the reactor trip signal and within 50 seconds bring the gas circulators up to their post-trip speed of 15 per cent. Figure 15 (lower curve) show reactoe sth r temperature transiene th r tfo condition in which only the gas circulators and the decay heat boiler in one of the four quadrants are assumed to operate. It can be seen that temperatures are well controlled.

An investigation of still more severe fault sequences with no restoratio post-trif no p coolin s showgha n that faultn ,i s where the reactor remains pressurised wilt ,i l take many hours before fuel cladding failure commences and, therefore, correspondingly longer befor y activitean y releas environmente th o et . Figure 15 also show lone sth g term trend fuef so l temperature sucn si h a sequence and it can be seen that the rate of fuel temperature

CLAD TEMPERATURE SAFETY LIMIT

FUEL CUD TEMPERATURE - °C

12345678 TIME HOURS PRESSURISER AG 5 1 FIGD. REACTOR FAUL LOS- T FEEF SO D

24 ris modess i e thatd tan , even afte hours0 r1 , fuel clad temperature lon a froy e gwa msar their safety limi 1,350"Cf to . Secondly depressurisatior ,fo n faults largese ,th t practicable pressure breath n ki e boundar coolans ga fractura e s yi th t n ei clean-up system. A complete severance of the largest pipe in this system diameterm (somc 8 e1 ) would fully dépressurise eth reactor in about 40 minutes. The reactor would trip from a low gas pressure signa llimitine andth n i , g design fault sequence, post-trip cooling is assumed to be provided by only two of the four quadrants utilising main boilers and gas circulators. The post-trip cooling system progressively speeds up the circulators for this fault to compensate for the falling gas density. Figure 16 shows the core temperatures intially rising significantly unti reactoe lth r trips. This ignore effece sth t of the auto control system in controlling temperatures at an earlier stage post-tri.e th Eve , nso p period temperaturen sca be wele seeb lo nt controlled.

In the same fault the consequences of not initiating any post- trip cooling have been explored. It was found, not surprisingly, that core temperatures increased more rapidly than in the equivalent pressurised reactor fault sequence but nonetheless Figurs ,a shows6 e1 rate ,increasf th eo n ei temperature is still modest. The fuel clad (melting) safety limit woul reachee db d only after abou hours3 t1 . However, with the reactor pressure lost the fuel pin internal pressures will cause the fuel cladding to creep and some failures would be expected about five hours after the fault. Even so, this

1100 -i

1000 - FIRST CLAD FAILURE

FUEL CLAD TEMPERATURC ° E-

MINIMUM POST TRIP COOLING

300

TIME - HOURS DEPRESSURISATIOR A6 FIG6 1 . N FAULT

25 represent considerablsa e perio timf o dwhicn i eoperatore hth s could act to restore cooling and, even assuming this conjunction of failures to remove decay heat, it is inconceivable that cooling restoree woulb t dno sucn i d timescaleha .

In the AGR, with on-load refuelling, it is necessary to consider the potential for creating a radiological release as a consequenc irradiatef eo d fuel being dropped durin removas it g l froreactore th m mechanise Th . m leadin suco gt accidenn ha t woul failurwhice r dsinglb e ba th he f eti o support fuee sth l elements nucleae Th . r safety implication sucf so evenn ha t have been thoroughly investigate supported dan y full-scaldb e stringer drop tests. It has been shown that, even at the maximum drop height intreactore th o automatie th , c trip facilities provided to deal with this situation would maintain the damaged fuel clad well belo mels wit t temperaturee Th . resulting off-site dose to the public would not approach the level at which evacuation procedures would be invoked and indeed i tprobabilitl woulal n i d y represent onl smalya l fractiof no this level.

To summarise briefly then, it has been shown that AGR performanc excellens i e t unde mose rth t extreme design basis fault conditions, but even more reassuring is the time available to institute remedial measures in those fault sequences, however improbable whicn i , h post-trip cooling system t operateno o sd .

Towards a New AGR

reactoe Th r desig Tornesr nfo s incorporates substantial margins. w cleano s rIi t that these marginexploitee b n sca increaso t d e the heat output from future AGRs.

A contrac produco t appropriat e desige th th e d nan e safety assessment of such a reactor has been placed with the National Nuclear Corporation (NNC) with finance from the CEGB, SSEB and NNC itself.

The first phase, confirming the main parameters shown in Figure 17, has been completed. This was aimed mainly at substantiating the higher output but was also necessary to examine any advance in safety criteria emerging froe Industry'th m s practice th r eo Nil's requirements oveyeare th r s sinc lase th et projects were launched. In particular, Nil specifically asked for a re- baffleassessmens ga e .beew th no f Thint o s completesha d dan provides a positive demonstration that the integrity is high, addition i s tolerant i bu t ni largo t e defects.

Work is continuing on the second, more detailed phase of the stud yproductioe aimeth t a dcomprehensiva f no e desigd nan safety submission.

Overview

Not surprisingl Chernobye th y l disaster brough immediatn ta e fal supporn i l nuclear fo t rU.Ke poweth .n i rDespit e explanations of the differences in design and operation of U.K. reactor differenced san licensinn i s g procedures politicae ,th l

26 PARAMETER UNITS HUNTERSTON B TORNESS PROPOSED (TYPICAL 1986) A.G.R. No. OF FUEL CHANNELS. 308 332 332 REACTOR HEAT MW 1535 1623 1740 MEAN CHANNEL RATING MW 4.98 4.89 5.24 PEAK CHANNEL RATING MW 6.44 6.26 6.70 GAS PRESSURE UNDER TOP SLAB bar a 39.3 41.0 42.5 CIRCULATO OUTLES GA R T TEMP. °C 2B6 298 297 BULK CHANNEL GAS OUTLET TEMP °C 634 639 639 CIRCULATOR POWER CONSUMPTION MW(e) 40.5 42.1 45.2 BOILER STEAM FLOW kg/s 515 52S 552 STEAM PRESSURE AT H.P.TURBINE bar a 155 167 160 STEAM TEMPERATUR H.P.TURBINT EA E °C 478 538 538 FEED TEMPERATURE °C 156 158 142 GROSS GENERATED POWER MW(e) 640 705 763 NET POWER SENT TOU MW(e) 592 645 700

R PARAMETERFIG.1 G A 7 S parties in oppostion are calling for either phasing out of nuclear power or a period of re-assessment before authorising new construction.

The report from the Inspector of the Sizewell Inquiry into the proposa expectes i CEGy immediate lb builR o th t B PW n i dda e future. Much now depends on the content of this report and how the Government, still strongly supporting nuclear power, decides to proceed. With the likely size of nuclear ordering programme, it would be unrealistic to build more than one nuclear system in the U.K. Even witmose hth t optimisti cnucleae vieth f wo r ordering programme, this would incur heavy penaltie costn i s s and performance in maintaining the infrastructure.

Clearly there would be no incentive to consider a change in the U.K. from well-tried CÛ2 cooled technology to embark on the lengthy busines bringinf gas-coolew so ne y gan d system sucs ha the HTR to commercial success. Notwithstanding the excellent safety characteristics of the AGR which have permitted siting clos urbao t e n locations beet ye seriouo n ,n s a ther s seha interest in combining these with heat production for large industrial processes. Nevertheless the continuance of gas-cooled technology in its present form in the UK must be of continuing assistance in the efforts to develop HTGR in Europe and the US, in the better appreciation of what gas-cooled technology has to offer, including the transfer of engineering experienc developmend ean t know-how, muc e whicb f ho n hca directly related to HTGR requirements.

27 HIGH TEMPERATURE GAS-COOLED REACTORS

R. SCHULTEN Institut für Reaktorentwicklung, Kernforschungsanlage Mich GmbH, Jülich, Federal Republi f Germanco y

Abstract

The historical development in energy technology goae th f realizinlo s ha g high temperature higd san h efficiencie energe th r y fo sconversio n processese Th . high temperature reactor (HTR) realizes this goal in the besy comparetwa otheo t d r reactor types uset i ; s ceramic material heae d heliuth an st s ma transfe r media e developmen.Th e fueth l f to element e th n o s basis of coated particles is considered to be finished. Future application productioe th e ar s f electricito n y productioe anth d procesf no s heaother fo t r areas of the energy market. Originally the thorium/uranium fuel cycl s beeeha n envisaged w enriche, lo toda e th yd uranium fuel f cyclinteresto s i e , because th f eo low prices of uranium. Typical for the HTR is its flexibility with respec differeno t t conceptse th n .O basis of the existing prototypes AVR, THTR-300 and Forth St.Vrai e followinth n developmeno tw g t linee ar s in discussion: reactors of small unit size and reactors of medium unit size e safet.Th y qualities have been demonstrate R reactoe operatioAV th e n y i b th dr f no Jülich operation i ,years9 w 1 whic no r s .fo ni h These experiences will be of great importance for future plants.

1. Fundamental concepts The historical goal of the technology of power station, as well as in energy technology in general, e realizatioith s d applicatioan n f higo n h temperatures processee foth r f energso y conversion e hig.Th h temperature reactor realizes this goal to the best compared

29 to other reactor 'types. With high temperatures in the range d betwee100C hig° an 0 h0 70 nefficiencie e sar realizee productioth r fo d f electricito n d othean y r application e energth n yi s marke e possibletar e .Th realizatio f higno h temperature possibls i s e because of the application of ceramic materials. In combination with requirements from the neutronic side the material for the moderator and for the construction of the HTR is graphite. As the coolant the inert gas heliu applieds i m e pressur .Th e primar th f o ey circuit lies between 40 and 60 bars. On the basis of these principles research, developmen d demonstratiotan n work has been done since about 25 years in the Federal Republic of Germany, in the United States, in the USSR, Great Britain and in Japan as well as in some other countries. e mosth tf o importan e On t successe e R+Dth - f o s work in HTR technology is the technique of the coated particles e Federath n ,i lFIG. 1 Republi. f Germano c y the pebble bed typ fuel element has been developed anuseds i de pebbl Th . e contain e coateth s d particles n inneia n r part, whic surroundes i h n outea y rb d layer fre f coateo e d particles. This concepa f o t fuel element is considered to be an industrial product.

Carùon layers

HTR fuel element

: HTFIG1 .R fuel element

30 Its development is also a success. An important featur s thai e t very high burn-up e achievedb n ca se Th . release of fission products during normal operation is very small e temperaturTh . e stabilit e fueth lf o y elements is also in the range of 1600 to 1800 °C very good. By experiments it has been shown that the release of fission products in this temperature range is minimal. Because of this good temperature stability the concepHTR-Modul-reactoe th f o t concepe th d tan r of the HTR-industrial-reactor have realized a particular safety concept. In this concept practically no radioactivity is released to the surroundings in the heaviest accident. In additio o that n t this concept realizes repairability after the heaviest accident. Federae Ith n l Republi Germanf o c pebble th y e type fuel elemen s beeha t n developen i s wela ds la the Soviet Union. In contrary to that the bloc type fuel element has been developed in the USA, FIG. 2. The main featur f thieo s fuel elemen hexagonaa s i t l bloc made from graphite. This bloc contains 'bore-holes

Blockförmiges Brennelement

1 Kühlkanal 2 Brennstoffstab 3 Abbrennbares Neutronengift 4 Bohrun Steuerstar gfü b 5 Greiferbohrung 6 Bohrung für Absorberkugeln

FIG. 2: Prismatic fuel element

31 fuee foth rl stick d bore-holesan guido t scoolane th e t helium. The bloc type fuel element is applicable for smaller and larger reactor cores.

2. Applications

The HTR is first of all, as well as other reactor types too, developed for the production of electricity. e presentlTh y established technolog o product , is ye e thaus to t stea d turbinen proe i man th - r fo s duction of mechanical and electrical energy. The HTR realizes conditions for that steam as they are used in conventional power plants. Fundamentall applie th y - catios turbinega f o n alss i s o possibl direcn i e t and indirect loops. This technology has been inten- sively develope e seventiethsth n i d . During this time the further realizatio s turbinega f o n se b coul t no d continued because the know how on the behaviour of the primary loop was not sufficient. The advantages of the application of gas turbines are remarkable, therefor s applicatio it ef interes o e e th b n y i tnma next decades. Beside the production of electricity the HTR is also applicabl productioe th r fo e f proceso n s heat. In cogeneration of electricity and process steam a broad variety of applications does exsist in chemical industry. An area of particular interest in the coming decadeenhancea productioe vi th l s i soi d f no oil recovery, here injection steam can be produced witHTRe th hn intensiv A . e research, developmend an t demonstration program has been done for the application of nuclear proces e refinemens th hear fo t d converan t - sion of coal, oil and gas for the production of environmentally benine liquid energy carriers, whicn ca h be use s fuemotoa r d fo lr heatin rfo car d san g purposes. The advantages are the following: application of nuclear energy instea f fossido e l th fuel r fo s process heat, reductio problemf no s with respeco t the environment, particularly the reduction of the

32 Problem carbof so n dioxide. Toda e latteth y r applications are not possible because of the low prices of expectes i t i t oild Bu .tha t these applications will be economically attractive in the coming decades and it is expected that they will be of great importance in future because liquid energy carrier e requiresar d

in the energy world supply and because the C02~effects climate oth n becomy ema eproblema .

3. Fuel cycle The fuel cycle of the HTR is very flexible. The original goal of the development work in the world has beeapplicatioe th n e thorium/uranium-fueth f no l cycle calculationn I . havt i s e been shown that with this fuel cycl eneaa r breederealizede b n rca n .I this conference there wil givee lb nreporta , which shows, that eve breedina n g procesrealizede b n sca . n calculationsOuow r , which have been finished five years ago, have shown tha conversiota n factof o r economicalls i abou 8 0. t optimume th y ,f o thi s i s course influence e reprocessin coste th th f so y b d f o g the fuel conversio.A n facto abouf o r8 realize 0. t s a utilizatio nucleae th f no r fuel whic s fivi h e times higher compare fuea o ldt cycle without reprocessing. Toda fuee th yl cycle witw enrichelo h d uranium withou) (% 8 t reprocessin witd gan h direct disposal is prefered. The reason is the low price of uranium and its sufficient availability. This fuel cycle has the advantage that almost all plutonium produced in the reactor is used. In addition to that the produced plutonium contains much higher isotopes; therefort i e dangerout ino s s with respec proliferationo t e .Th utilization of uranium with this fuel cycle is almost as high as in the fuel cycle of a .light water reactor (LWR) witreprocessine hon d recyclin gan proe th - f o g duced plutonium. In contrary to this fuel cycle pro- Federae poseth n i dl Republi Germany,f co ,fuea l cycle with an enrichment of 20 % will be used in the united States.

33 R conceptuaHT 4. l designs e pebblTh e type fuel element allow higa s h flexibility with respect to the dimensions of the reactor possibls corei t I . o realizt e e reactor cores with small diameters and relatively large heights, as well as cores wit a largh e diamete d smallean r r heights. e principlTh e pebblth d reactof o be e s explainei r d in FIG. 3.

top reflector p shielto d graphite guide for shutdown rods

biological — shield

liner — side reflector thermal — reactor core shield _ outer vessel reflector — inner vessel bottom fuel element discharge pipe support structure

principle Th pebbl e : th 3 FIGd reacto f .be e o r

The pebble bed is cooled by the heat transfer medium helium e pebbl.Th d typbe ee fuel elemente ar s transported by a pneumatic system to the upper part of the core, and they are extracted from the lower part n principlI . o fuetw e l loading mode e possiblear s : there is the once through than out mode and the multiple mode whicn i ,e fue th hl element e recirculatedar s . Both fuel loading modes have been investigated inten- sively. During this conference the differences of the two modes will be discussed.

n additionaA l flexibilit realizes i y d because o typetw f reactoo s r pressure vessel e feasiblesar :

34 the prestressed concrete e steevesseth ld lan vessel . The prototype for HTRs with steel vessel is the AVR in Julien. This reacto s durinr ha s almosit g year0 2 t s of operation resulted in fundamental know-how on the HTR, FIG. 4.

/ Sltam Header 2 Containment 3 Outer Pressure Vessel 4 Inner Pressure Vessel 5 Steam Generator 6 Quench Tank 7 Upper Reflector 0 First Biological Shield 9 Pebble bed Core 10 Lateral Bottomand Reflector 11 Fuel Oischarge Pipe 12 Gas Purification System 13 Dismantling Machinery 14 Second Biological Shield Î5 Charge room Fuelfor Elements K Cooling CirculatorsGas 17 Dismantling Machinery Circulatorsfor 18 Components of Fuel-handling System 19 Material Lock 20 Heavy Load Elevator 21 Circular Bottom Support 22 Storage Channel for Spent Fuel Elements

FIG. 4:

AVR-reacto n Jiilici r h

Germae Th n reactor constructor basise th hav sn o e of the AVR developed interesting concepts for small HTR units t firs.A woulI t d lik mentioo t e Module th n - e companth f o y R InteratomHT , FIGn additioI .. 5 n to tha woulI t d lik mentioo t e HTR-100-concepe th n t of the company HRB, FIG. 6. Both proposals are presented

35 FIGModul-reacto: 5 . r (INTERATOM)

+ 42,0

+ 12,5

1 Spent fuel building 2 Reactor auxiliary building 3 Reactor building 4 Residual heat removal building 5 Turbine hall 6 Industrial steam plant

FIGHTR-100-industria: 6 . l reactor (HRB)

in this conference. It is my opinion that these small f interesR o unitl application HT al e sar r fo t n i s which smaller power f interesto rate e e th sar y .B combination of several small units power plants in the A e realized b US 0 MWe y e 50 lma th o t n I .0 rang20 f o e e ide f th smalleo a o to r reactor discusseds i s . This interesting development will also be presented in this conference.

36 The second concept of the HTR is characterized by the prestressed concrete vessel prototyp.A f thieo s THTR-30e th kin s i d Schmehausenn 0i . This reactor went into operation during this year and has reached 100 % electricity production recently, FIG. 7. This nexe th reactot base r th HTR fo eHTR-500e s i rth , . The planning of the HTR, which is done by HRB and a

nucleae MW 0 r30 FIG powe: 7 . r plant Hamm-Uentrop

grou f utilitiepo s startesha d recently n thi.O s conference there will be report about it. Parallel to thimentionee b so t ther e Fors th deha t StVrain-reactor in the USA. This reactor is in operation since years, about the result a report will be given during this conference.

5. Safety qualities The AVR-reactor in Julien is excellently suitable to demonstrate various safety qualities of the HTR. In particular it has been proofen very good as a plant for the tests during the development for fuel elements.

37 operateR AV meaa e t Th nsa helium outlet temperature . Eve °C t thia n0 so95 fhig h temperatur e primareth y coolan relativels i t y clean e contaminatioTh . s i n very low. In the AVR-reactor experiments have been performe coolane whicn i dth e los th f ho s t without scra s beemha n simulated e result.Th s have been that the negative temperature coefficient of the reactivity is fully effective and that a stabilization of the temperature niveaureactoe th n i sr core takes place. o inherent e du Thi s ti s transport mechanism r heafo s t in the core. In 1978 there happend an incident by an ingress of water from the steam generator. This incident has produced much know how on effects on steam and water to the reactor core. The operation R wilAV le continuoth f e yearth n si e 198 d 19887an . The main purposes are experiments which are safety related. The concept of the small HTRs in the range of 200 to 250 MWth has been developed on the basis of the experiences AVRe gaine th e mos .n Th i dt importan- as t e s conceptha i e smalth pecn heav i th tR f o tf lHT y o t accident e maximuth s m temperatur e corth e f o edoe s not exceed 1600 to 1800 °C in the reactor core. Because of this the release of fission products from the fuel elements is rather small. Therefore these reactors have a particular safety quality. In addition to that there is a fundamental possibility of repairability after heavy accidents. It is my opinion that this reac- r concepto t becauss particulait f o e r safety quality is of great importance for the future. As mentioned before there are the plans from the utilitie d fro e companan so realiz t th m B HR y a efollowe r - reactor to the THTR with a power of 500 to 550 MWe. In this concept particular safety qualitiee th f o s concrete vessel wil usede lb .n interestina Righ w no t g experimental work is done in the KFA Jiilich on the behaviou concretf o r e vessels o demont goae s .Th i l- strate heat resistanc concrete th f eo e upto temperature of 1300 °C.

38 6. developmene Statuth f o s t researche e statuth Th f so , developmen demond an t - summarizes stratioi R HT e nfollowss a dth wor n o k : developmene th - pebble th f eo t type fuel elemend tan the bloc type fuel element is almost finished. The high quality of the fuel elements hase been demonstrated by the operation of plants as well as by large experimental programs. - There have been developed concepts for small HTR on th e operationa th bai f o s l experience e AVRth -f o s reactor e Federath n I .l Republi f Germano c wels a yn i l Unitee ith n d States ther e undertakeear n efforts for the realization of such small reactors and for the demonstration of their particular safety qualities. The planned small reactors in the USSR also belong thio t s s category. - And finally with the successful operation with the THTR-30 might i 0 e possibltb o offet e f followero r - reactor with medium unit size.

39 EXPERIENCE WITH GAS-COOLED REACTORS (Session B) DEVELOPMEN HIGA F THO OUTPUR TAG

B.A. KEEN National Nuclear Corporation Ltd, Knutsford, United Kingdom

Abstract

papee Th r R outlineAG w designe e sa th f no Heyshae baseth n do m II/Torness reactors whice har nearing completio theif no r construction stagee Th . intent is to replicate as far as possible the Heysham II/Torness desig meetinn ni targee gth t station generator output of 1526 MW(e), and this can be achieved by design changes only to the HP steam pipework. Two additional design options, a vented containment and a dry refuelling route, are also under consideration.

1. INTRODUCTION The Advanced Gas-Cooled Reactor Power Stations now being constructed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland, represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station MW(e0 66 outpux ) 2 grossf to , althougs i t i h currently intende uprato dt possibls a er fa thi es sa towards the figures discussed later in this paper. The construction of these new AGRs has been to programm withid ean n budget. Fuel e loadinth r fo g first reactor at Heysham is expected late in 1986 wit othee hth r three reactors following ovee rth subsequent twelve months desige Th .botf no h station bees sha n successfue baseth n o d l operating AGR Hinklet sa y Poin Hunterstod tan n which havw eno bee servicn i n almosr efo yearsn tte , although minor changes were made to meet new safety requirements and to make improvements suggeste operatiny db g experience. Sizewele th t 1 PubliA 'B l c Inquiry CEGe ,th B gav commitmenea s optionR i o retaitt AG C e NN n.th engagedefinitioe th n e improvei dth f nR o dAG design. In order to use the invaluable experience gained stationR sinc AG lase o eth t tw s were ordered, it was decided that the new designs should be developed from the Heysham II/Torness design. As part of this study, it was necessary to address the

43 specific safety requirement futurr sfo e AGRs R InquirNucleae presentePW th e y yb th r o dt Installations Inspectorate. In the interest of continued safety improvement aimed at mitigating the consequences of accidents however unlikely these may be, consideration is also being given to a design of vented containment. The move by the CEGB to construct a national dry fuel stor creates eha d refuelliny interesdr a n ti g route as another possible development shoult I . notee db d that the design study is not yet complete and no decisions have yet been taken on these options.

2. REACTOR DESIGN desigThw ene powef no r statioo tw s nha identical reactor 174f so 0thermaW M l output driving individual turbo-generators givin statioga n gross electrica MW(e)3 ladditionae 76 Th outpux . 2 f to l turbing/generatorratine th f go s wil accommodatee lb d by small changes to the design of the Heysham II/Torness units. Principal dimensions and design parameters are given in Table 1. The reactor is contained in a single cavity prestressed concrete pressure vessel (PCPV),a section through the vessel being shown in Figure 1. e reactoTh r cor supportes i e diagria n do d an d enveloped by the gas baffle, which is welded to the pressure vessel liner floo preveno rt t movemenn i t the event of an earthquake. The gas baffle enables the graphite moderato othed ran r core structural components to be adequately cooled during normal operation. reactoe Th r fue mads i l e from slightly enriched uraniue for sinteref th o m n i m d uranium oxide pellets pellete Th . assemblee sar multi-starn i d t spirally ribbed stainless steel cans to form fuel pins, 36 of which are enclosed in a graphite sleeve to form a fuel element. The fuel elements are to the Stagw ne desigI eI n with single thick sleeveo st improve their resistanc fueo et l handling loadings, and wit loweha r flow resistance. Each fuel assembly comprise bottosa m support uni reflectord tan , eight fuel elements, a top reflector and a lower gag unit threaded onto a tie-bar which is suspended from a plug unit which incorporates a flow control gag unit. Eight gas circulators deliver carbon dioxide gas at 42.5 bar a and 297°C into the plenum below the diagrid. Heat is transferred to the gas from the fuel raising its temperature to 645°C. Main boiler units are arranged circumferentially in the annulus formed by the gas baffle and the reactor pressure vessel liner. Each boiler has high pressure and

44 TABLE I PRINCIPAL DIMENSIONS AND PARAMETERS

Pressure Vessel Internal height 21.9m Internal diameter 20.3m Wall thickness 5..8m slap To b centre thickness 5. 4m Bottom slab centre thickness 7.5m

Reactor Primary Circuit baffls Ga e internal diameter 13.9m Gas baffle dome height above liner floor 19.9m Core restraint tank outside diameter 13.6m Typ maif eo n boiler 2 start, once through Serpentine platen rectangular unit Numbe maif o r n boilers 4 Number of boiler units per main boiler 3 Typ decaf eo y heat boiler 1 start, once through Serpentine platen Numbes circulatorga f o r s 8 (2/quadrant) Type of gas circulators Centrifugal, single stage encapsulated, varaible inlet guide vanes Reactor Core Numbe f fuero l assemblies 332 Number of control assemblies 89 Numbe f secondarro y shutdown system channels 163 Heigh corf to e (including neutron 12.8m shields) Overall widt corf ho e across flats 12.4m (regular 16 sided polygon)

Reactor Parameters Reactor heat 1740 MW Bulk circulator outle temperaturs tga e 297°C Bulk fuel channel outlet gas 645°C at top of fuel temperature stack Net circulator flow 4697 kg/s Gas pressur circulatot ea r outlet a r 45.ba 2 Pressure difference acros baffls sga e dome r 2.5ba 6 Total primary circuit pressure drop 3.02 bar Peak channel power 6.7 MW

Boiler Parameters Feedwater flow 552 kg/s Feedwater temperature 142°C HP steam temperature 538°C steaP H m pressure a r ba 0 16 Reheater steam temperature 539°C Reheater steam pressure 40.7 bar

Turbin Generatoe- r Parameters Turbine heat rate 2.31 kJ/kJ(e) Nominal gross electrical output/reactor 763 MW Nomina electricat lne l outpu statior tfo n 1400 MW Design lifetime 30 full power years

45 TYPICAL FUCl SWHOPIPE — —COHTHOL SIANOPIPE

TOP MAN ACCESS PENETRATION

•PRESSURE VESSEL

BOI REHEAR L£ T MLET -^ AND OUTLET PEKEmitON*

-REHEAT CASINO

SECONDARY SMUIDOW« ROOK

FIG 1 REACTOR PRIMARY CIRCUIT SECTIONAL ARRANGEMENT OF REACTOR UNIT BASIC VAULT reheat sections with separate feed and steam penetrations through the outer cylindrical surface of pressure th e vessel maie Th .n boilers ga d san circulator physicalle sar y divided into four separate quadrants by division plates at the circulator inlet plenum below the boiler seal. During operation, feed water is supplied to the boilers at 142°C and 217 bar a giving steam temperatures ga 538°Cf so e Th . circulators are driven directly from grid supplies usin froW e statioM mth 5 g4 n output s floGa .w control is achieved by adjusting the angle of guide vanes at the inlets of each circulator. A totally diverse secondary system for the removal of decay heat is provided via the decay heat boilers (DHB). These are commissioned automatically following a trip and have their own feed system and heat sink. primare Th y syste contror fo m shutdowd lan f no the reactor comprises absorber roddrived san s housed in standpipes in the top cap of the PCPV. The actuator and chain store is mounted between a closure unit and radiological shield plug through which the chain passes and from which hang the articulated tubular control rods.

46 A secondary shutdown system (SSD) designeo dt inject nitrogen into the core is provided as a diverse mean shuttinf so reactoe gth re dowth n ni remote fault situation, should a predetermined number of control rods fai enteo lt r intcore oth e when required. This system operates from the SSD room beneat PCPVe hth . Boron glass injecte e beadb n sca d operatoe byth r intcore provido th e t elong-tera m reactivity hold-down capability in the event of continued failure to release the primary shutdown rods befor reactoe eth depressuriseds ri . Both reactors are served by the fuel handling facilities located between them, a single fuelling machin othed ean r shared equipment. These facilities providstorage th r assembld efo w fuelan ne e ,f th yo reactore th f o decad transfest an ou y d an f fue ro n i l storage tubes, and subsequent dismantling of the fuel, servicin reusablf go e plug unit componentd san disposa debrisf lo subsequene ,th t pond storagf eo irradiated fuel and finally its despatch from site.

3. DESIGN CONSTRAINTS maie Th n constrain developmene th n to e th f to desigreplicato t s nwa closels ea possibls ya e eth Heysham/Torness design. Only change meeo sw t tne safety requirement simplificatior so desige th f no which woul leat significano dno t d t alterationo st station layout were considered. Operating constraints were also applied to ensure that the experience of the existing AGRs were still valid and that the safety case prepared for Heysham/Torness remained applicable. Within these constraints, and with the exceptions given beebelows ha n t ,showi n that sufficient flexibility remains in the design parameters to allow for uncertainties and to ensure that the target station generator output of 1526 MW(e) wil achievee operatine lb th r fo d g lifetimf eo the plant of the equivalent of 30 full power years. It is possible that greater output might be achieved after commissioning has demonstrated the expected available margins.

4 HAZARD DESIGN 4.1 Earthquakes A new enveloping Safe Shutdown Earthquake (SSE) has been defined for all future nuclear stations in base s i . Thi historican UK E do se SS th l evidencf eo seismic activit Britise th n yo h a mainlans ha d dan peak acceleration of 0.25g. The effect of this earthquake on the buildings and plant will be dependan soie th l n to structurE SS site e e th Th .t e a loadings have been derived for the two extremes of UK

47 site, with the most onerous conditions being used as desige th n basis essential Al . l plan structured tan s are designed to survive this design basis SSE without preventin shutdowe gth post-trid nan p cooline th f go plant orden I . preveno rt suddeta n losf so capabilit earthquaket ya s with accelerations just beyon desige desige th dth s f o nha n m basisai e ,th bee provido nt conservativea E e SS respons e th t ea such that adequate shutdown and cooling would be availabl earthquaker efo 0.35o t beyond d p sgu an . 2 Aircraf4. t crash The frequency of aircraft crash on the vulnerable statioe areath f bees so nha n assesser fo d typical sites "outside" and "within" Areas of Intense Aerial Activity stude Th .y confirme e impacdth t frequency was very low for both areas and that with the high degree of separation and segregation of plant in the Heysham II/Torness layout, the risk of causing a radiological hazard was not significant. additioe Th fira f eno fighting reactoe systeth n i mr pil arep contaieo ca at n aviation fuel fires reduces the risk further.

5. DESIGN CHANGES The effect of the higher output on the design of Heysham and Torness has necessitated significant design changes only to the HP steam pipework to accommodat highee eth r steam flows without excessive pressure dropsmeeo extendet e td th ,an d station lifetime. Manufacturin constructiod gan n experience gained durin Heyshae gth Tornesd man s projects sha been used to make detailed changes in various areas desige ofth simplifo nt y manufacture, ease construction and/o reduco rt e costs. e Nonth f eo changes adopted will significantly affect the object of replication. The safety cas on-loar efo d refuellint ga HeyshaTornesd an onlI s I msha y been presenter fo d refuelling in batches as part power levels with consequential loss of output. Simplification of the protection systems will enabl time eth e taker nfo each reducedbatce b o ht , while further generic work on refuellin expectes gi provido dt meane eth r sfo increasin presene gth t limi refuellinn to g power towards full power and single channel refuelling. These improvements will giv significanea t increase in reactor availability.

48 6. VENTED CONTAINMENT AGR and Magnox Safety Cases have always been basie madth sn eo tha crediblo tn e accidene b n tca identified which would require secondary containment reactore ofth . Until recently Heyshar ,fo m II/Tornes earlied san r AGRs probabilite ,th a f yo dropped fuel assembly during refuelling had been considered to be so low that it was not necessary to taksafete th e n yi accoun t casesi f to . However, shoul occurt i d , vented containmen sub-pile th f to e cap volume would reduce the activity release from such an accident. Other potential release areas in the reactor building have also been considered, even thougo hn accidents have been postulated which would require such provision areay ke se Th identifie. 2 C0 r fo d discharges are: Pile Cap Quadrants RooD SS m C02 Bypass and Clean-up Circuit Fuel Decay Store five Eacth e f ho area s wil hign l ow hhav s eit integrity ventilation system, broadly based upon those which currently exist on Heysham II/Torness AGRs. evene pressurbreace Ia nth th f to f ho e circuit, the normally running ventilation plant is isolated and a hot gas release route is opened to discharg escapine eth directls gga atmosphereo yt . The new plant connected to the above ventilation routes by normally closed dampers, would comprise heat exchanger, high efficiency particulate charcoad an r ai l filters extracd ,an t fans dischargin stackw planne e a Th . t o g t woul d starp tu automaticall hign yo h temperatur2 C0 r eo concentratio whe d discharge nan nth e flow througe hth hot gas route reduces to the capacity of the new plant, dampers would closdischarge routth o e t l eal e through the new plant. For discharges within the capacit filtee th f ryo system, isolation woule db immediate. In the unlikely event of a dropped assembly, fuel damag activitd ean y release woul immediatee db , therefor planw ne te ewoulth d considerably reduce eth consequences. Other breaches within the design basis, although initially exceeding the capacity of the plant, would not give rise to early fuel failures. For faults beyond the design basis, the plant would mitigate the consequences by reducing the total release of activity.

49 7. DRY FUEL ROUTE Recently the CEGB have decided to build a large central dry store away from the reactors, capable of accepting irradiated fueactind buffea lan s ga r betwee e finanth theld an mreprocessin t ga Sellafield. Thi developmenw sne t make logicat si o lt review the possibility of adopting a dry fuel route y futuran r efo AGR. The present sequence at Heysham/Torness is for irradiate buffe0 3 f o r d e place e b fueon o n lt i d storage tubes after removal froreactoe e th mth y rb fuelling machine. The buffer storage tube is water cooled, the heat being transferred from the fuel stringer to the water by natural circulation of pressurised CC>2. Afte day8 2 re deca sth y heas tha reache levea d l wher atmospherit ea coolinr ai y gb c pressure is adequate and the fuel is transferred to o Irradiatetw onf o e d Fuel Disposal Celle b o st dismantled. The elements are then transferred to the pond wherstores i abour t i efo 0 dayd 10 t s before dispatch. routy dr e woulTh d follo e samwth e sequence, but the number of buffer storage tubes would need to be increased to 50 to allow storage for 100 days. The increased decay time is required to permit the decay hea reduco t acceptablo et e levels before eth story dr fue ee places i lafteth n i dr passing through storey dr e fuee , th th l D cell IF n wilI .e l th remain fo minimu ra day5 despatche26 e s b f beforn mo ca t ei d fro y fuemdr la sit flaskn ei . The dry store is separate from the reactor building with loadin unloadind gan d en g e cellon t sa so that the store may be extended during station life should this become necessary initiae Th . l capacity proposed is 1280 elements, stacked 16 high in 80 tubes, cooled externally by natural circulation of air. There are no technical obstacles and a dry route would remov doubty ean s abou long-tere tth m storag fueR AG l f eo afte r passing throug ponha d system, although development of a dry road transport flask woul requirede db .

8 CONSTRUCTION PROGRAMME experience Baseth n do e gaine Heyshat da d an m Torness the construction period from start of permanent work commerciao st lfirse th loa tf do reactor woul reducee db month3 months5 7 y db o st .

50 9 LICENSING PROGRESS As the high output AGR is very similar to Heysha Tornessd man , their Station Safety Report, SSR, licensine form th basie y futursr th an fo sf geo AGR in the UK. The SSR was submitted to the Nil in 198 preparation i 6 start-ur firsnfo e th tf po reactor at Heysham assessmenn A . extre th af t o requirement s raised by the Nil at the Sizewell Inquiry has shown that while extra safety studies will be required, no significant design change expectee sar madee b o .dt

10 CONCLUSIONS The successful progress of the Heysham and Torness projects in the UK coupled with the good operating experience of the similar Hinkley Point and Hunterston reactors form stronsa e g th basr efo design of a high output AGR. In producing the new design, the major aim has been to replicate the Heysham and Torness AGRs with the minimum of changes e increas o 152Th .t % 6outpun 15 ei f to MW(e) gross generator output coupled with a reduction constructioe inth n programm costd ean s improvee sth competitiveness of the new design. Wit modesa h t investment furtheo ,tw r developments, vented containmen refuelliny dr a d tan g route, coul addede d b formee Th . r would mitigate eth release from faults which are beyond the present design basis e latteTh . r woullogicaa e b d l sten i p view of the intention to build a national dry fuel store. An assessment of the Nil requirements for a Sizewele states a th K t U da e l th Publin futuri R ceAG Inquir shows yha n tha majoo tn r difficultiee sar expecte obtaininn i d operatinn a g g license, although design substantiation will be required.

51 TECHNICAL EVOLUTIO OPERATIOD NAN F NO FRENCH CO2 COOLED REACTORS (UNGG)

Y. BERTfflON CEA, Centre d'études nucléaire Saclaye sd , Gif-sur-Yvette, France

Abstract

e technicaTh l evolutio five th e f nFrenco £ coolehC0 d reactors (UNGG) from 1981 to 1986 needs to be outlined. These technical evolutions concerne e fueth dl elemen f Bugeto whic1 y h is now slightly enriched, as well as the load reduction operation required by the grid. In addition work in underway to increase the safety at the two St Laurent units, or to repair the hot steel upper-structures of Chinon-3 unit.

1. INTRODUCTION e technicaTh l evolutio five th e f Frencno 2 coolehC0 d reactors (UNGG) from 1981 to 1986 needs to be outlined. This evolution will be illustrated by examples taken from a survey made at each nuclear power plant. I will speak of the steel corrosio repaire th d snan (ISISChinon-e th t )a 3 unite ,th erosio corrosion- heaf no t exchangeo tw e d safetth ran t a y St Laurent units, thé graphite corrosion and the load-reduction operatio Bugeyt na . Evolutio e lasth tn ni year s illustratei s y figurb d whic1 e h give cumulatee sth d load factor f eac so predictee hth uni d tan d cumulated load factor quoted PEON. One can see the shutdown for repair.at Chinon-3 unit started on May 4, 1984. But figure 1 does

i — l — i — l — l l — i — i — i — l — \ — i — i — i — i nb.d'annee 1 2 34 56 7 8 9 10 11 12 13 14 15 16 17 18 18 20

FIGURE 1 : UNGG cumulated load factors versus time

53 not give a correct idea of the availability factor of the last predominance twth o year, on fro; sw nucleaf mno eo r generation in the French electrical network will require the nuclear units to take part in load-following and frequency adjustment of the gridR unitn FrancPW I .e s th eequipe d witcontroe th h l modeG allowing power variations are cheaper than UNGG units. Also the grid requires load reductio UNGe th G f o reactorsn .

E CHINOTH 2. N NUCLEAR POWER PLANT Three UNGG nuclear reactor e locatesar Chinonn i d . . The CHINON-A1 reactor which shutdown definitively in 1973 now serves as a museum and this since February 3, 1986. . The CHINON-A2 reactor was shutdown for economical reasons in June 1985 afte 0 year2 r operationn i s s cumulativIt . e load facto s 73.wa r 7% i.e .PEOe closth N o t predictioe n (figur, 1) e and have a last run of 869 days without scram. The management of the fuel load was optimized and during the last 8 months no new fuel elemen s loadedtwa e radia.Th l shufflin f fueo g l elements e overmoderatioanth d n obtained wit e feedine fueth hth l f o g chanel by graphite log provided the needed reactivity to operate and produced 970 GWeh of extra electricity. The average availability factor of the last five years was 86 '% with a maximum of 94.5 % in 1984. During the life of this reactor, 133,374 fuel elements were loaded with onl cla6 y d failures. Some shuffled t cell fuer examinationho fo sn l i elementw no e .ar s Good results are expected and the radial shuffling will be planned for the end-of-life of other UNGG reactors in order to save fuel. Table I show reliabilite th s f UNGo y G fuel elements.

TABL . I RELIABILITE F FUEO Y L ELEMENTS (30/06/86)

Graphite Core fuel elements Annular fuel element Numbef o r SICRAL F1 SICRA1 F L

. unloaded fuel 369,020" 75,550 elements . loaded fuel 542$250 88,140 elements . clad 10 3 failures

failure rate 2 10~< 5 < 4 10~5

CHINON-é Th . 3 reacto bees rha n shutdow n, 198 4 sinc y 4 eMa in orde repaio t r e steerth l corroded upper structure th f eo reactor. Figure 2 shows a vertical section of this unit. The corrosion did not have the same consequences on the Chinon-A2 reactor whose structures were bolted instea weldef do d (figur, 3) e e subsequenth n o r tno reactors 2 becaussteeCÛ le th ecorrosio n was already known. This operation of structure repairing is called ISIS and the cost up to mid-1987 will amount to 300 MF. The

54 Tbntde manuhpnhon 15 r.

feiton

FIGURE 2 : CHINON-3 reactor vertical section

CHINON 2 CHINON 3 (20 years)

Oxide formation

>. FAILURE WITHI 0 YEAR1 N S

Bolted Welded

Hot screak temperature ; FIGURE 3 : Mild steels corrosion mechanisms by C02

erection of a building was necessary to shelter a true-scale model of a sixth of the upper structure. Rigorous correspondence betwee uppee th modee rnth d an lstructur s i e obtaine a lase y b dr telemetr reactore y th don n o e . Five robots are necessary : three on the reactor, i.e., one for the camera, metae weldede o b th tw o t lr e fo .Th e on e tool th d an sr fo e on others are on the model for learning purposes. The reactor will operate agai mid-1987n i n . Steel corrosio s responsiblnwa r fo e

55 the nominal powee (seMW r insteae 0 lowerinMW 48 0 f o d36 : g Tabl eothee II)th rn .O reactors nominae th , l powe s reducewa r d for different reasons t loca.Bu l defects where steel corrosion is involved have been noted and their evolution is being carefully watched.

TABL . POWEII E F UNGO R G REACTOR

Design Power Effective Power UNIT Difficulty (MWe) (MWe)

CHINO2 NA- CHINO3 NA- 480 360 Stell corrosion SL1 480 390 [ erosion-corrosion SL2 515 515 on heat exchanger Bugey 540 540 graphite corrosion

3. THE BUGEY NUCLEAR POWER PLANT UNGA G uni locates Bugee i t th yn o dnuclea r power plant. This unit differs from the others in its fuel elements which are externall d internallan y 2 pressuryC0 e cooledth en i wels ,a s a l bar3 whic4 s i hinstea bars9 2 f .o d Thes o factortw e s allor fo w a greater specific power higt .Bu h specific powe higd an rh pressure mean high rate of graphite corrosion. The maximum local graphite corrosion measure 22.s 8.5t i da 4% 8 aepp (year equivalent full power) mechanicae .Th l behavio f suco r corrodea h d blocs i k poor, and mechanical and seismic calculations are under way in orde o evaluatt r maximua e m acceptable leve f corrosioo l o t d an n determine the remaining lifetime of the reactor. To day the lifetime is estimated to be 2 aepp. So this unit is now considered as an energy tank. In France the marginal cost of PWR is the chea- pest and these reactors operate in load following mode. So from an economical poin vief o tgrie wth d necessitate loaa s d reductio f thesno e UNGG reactors. Therefore e optimath , l management of this energy tank prompt operatoe sth defino t r e several operation levels acceptablreactoe th fuee d th ran l y (Tableb e III).

TABLE III. OPERATION LEVEL

Power (MWe) Comment

540 Exceptional grid demand 470 No more tha0 hours/yean50 r 400, 3000 ,20 Normal operation level upon grid demand Ten times a year.

56 This operating mode requires more from the fuel elements, so in order to increase the load-reduction number the fuel behavio examines i rt cells ho n o reduci d.T e graphitth e e corrosion, methane is injected into the CÛ2 with a 450 VPM level. The decompositio methanf no e produce carboxyhydrogenatea s d deposit which reduces the heat transfer from the fuel, and the reactivity with the neutron capture on hydrogen. The high hydrogen content in the stac mads slightla kha e f o necessar e yus enrichee th y d uranium (0.76 % instead of 0.72 %) since June 28, 1984. This enriched fuel gives better power distribution, and flexibility. The low increase in cost is compensated for by saving fuel in the outer e reactorth par f o to increas.T e outeburnue th th e rf o p fuel radial shuffling is not used at the moment.

E SAINT-LAURENTH 4. T NUCLEAR POWER PLANT o UNG e locateTw Gar unit 2 n thio dSL ssd callean site 1 .SL d Figur show4 e verticaa s reactor1 unil2 SL SL sectiote e th . Th f no is very Vandelloe similarth s i s ,a s uni n Spaini t e nomina.Th l power was reduced (Table II) because of a erosion-corrosion problem of the heat exchanger. This phenomenon originated from a concep- tion flaw, and the vaporization level was changed in reducing the power. Meanwhile, during the first operating years, the exchanger s affectewa y erosion-corrosionb d todad ,an y some leaks still occur.

The heat exchangers are divided into two separate parts and the leaks affected mainly one of the two halves in both SL1 and Vandellos , somSo .e other phenomeno y als nma e involved ob e .Th water quality is important, and at Vandellos better results were recently obtained with AMP than with morpholine. Over the last five years, however, work has been focussed on a réexaminâtion of the Safety Report in the light of feedback experience. Among these works are : . EAR use as ultimate emergency system. buildine Th n indépendan a . f o g t emergency control room named BUS. . Increasing blowing availabilit reduciny b y commone th g - causes failure with fire retarding bulkheads and the separation of cableway, and by supplying auxiliary sets with main steam. . The possibility, in case of scram, of feeding turboblower witw qualitlo h y main steam. This forced running provides more than hal houn a fblowingf o r . falsf o e e us tulip-shapee Th . d fitting eacn o s h fu.el channel in order to prevent plugging. In addition to the erosion-corrosion problem and work on safety, a third point concerning the two Saint-Laurent units is a thirty-year extension of their life. To "achieve this extension, the consequence f steeso d graphitan l e corrosion will e havb o t e studied.

* heat exchanger cooling the outage unit.

57 ENSEMBLE DU RÉACTEUR Coupe verticale

28.500 J32.300 ^131.100

câble8 13 s

92.250

Nota : la (urbo-soufflante est ramenée dans le plan de coupe. Tension appliqué« a chaque câble 185 t. elfe. Longueur total cibles ede s 191.5km Poid métru sa chaque éd e cible 12.g 3k Ceint, sup. + ceint, inf. 11,4 km Dalle sup + dall. e inf. 37,m 8k

FIGUR Saint-Lauren: 4 E reacto1 t r vertical section

58 TABLE IV. HEAT EXCHANGER LEAK NUMBER

UNIT leak number SL1 39 SL2 15 Vandellos 82

5. CONCLUSION Nuclear UNGG power plants toda e somewhaar y t overshadowed reactorR bPW y s which Francen i , , producl e al mor f o e % tha 5 6 n electricity generated at the lowest cost. Nevertheless, there is reason to be proud of the fact that these UNGG reactors, with which France entere nucleae th d r age, have aged well despite their early flaws and discontinued commercial development. These flaws, stee d graphitan l e corrosion, erosion-corrosion of heat exchangers, were compensated by an acceptable reductio nominae th f o nl power. ISIS work showee th d realism of a large-scale undertaking on upper reactor internals. The Chernobyl event has widened the field of our safety research on the UNGG reactor type. Indeed, we now need to finct the righ a severn cast i f thingo eo d e o accidentt s r instance.Fo , the use of lead would be ill-advised because it would create an eutectic with the magnesium fuel clads.

59 FORT ST. VRAIN PERFORMANCE

H.L. BREY Public Service Company of Colorado, Denver, Colorado, United State f Americso a

Abstract

Fort St. Vrain, on the system of Public Service Company of Colorado, is the only high temperature gas-cooled power reactor e Uniteth in d States. This plan s designewa t y Generab d l Atomic Company and utilizes helium as the primary coolant. The core consist f triso-coateo s d uraniu d thoriuan m m fuel particles cast into cylinical rods within prismatic graphite blocks. The primary coolant system e consistreactorth f o s , twelve steam generator modules, and four helium circulators contained within a prestressed concrete reactor vessel. The once-through steam generators provide high quality 1000°F, 2400 psi main steam and 1000°F reheat steam to the main turbine. The plant has generated 3,334,000 HWH of electricity and has undergone 3 reactor refuel ings. Many of the primary system components at Fort St. Vrain are proto-typica n naturi ld thian e s plan s undergonha t n extensiva e e and elaborate testing program. Difficulties primarily with the helium circulators and fuel element column movements particularly t higa h power levels have resulte n significani d t modifications. Power operatio t 100a n s achieve%wa n lati d e 198t 1 witne a h thermal efficienc f nearlo y y 39%. Excellent performance th f o e coated fuel particles has contributed to a full power circulating activity sixty times lower e thadesigth n n with subsequent cleanliness throughout the plant. Except for 1985, when a combined total radiological exposure for all personnel reached 35 manrem due to the necessity to completely refurbish all of the control d drivro e mecanisms e averagth , e total combined exposur r eacfo e h e pasth t f o five years undewa s3 manremr . This exemplary radiological cleanliness has, with the exception of tritium, allowed For. VraiSt to consistentlt n y operate with nobl s airbornga e e and liquid effluent releases more than an order of magnitude lower than the average for the U. S. nuclear power industry.

For. VraiSt t n e syste(FSV)th f Publin o mo , c Service Company of Colorado, is the only high temperature gas-cooled (HTGR) power e reactoUniteth n i dr States. 0 Thimegawat33 s t electrine t c generating plant feature a helium-cooles d reactor witn a h uranium-thorium fuel cycle. PLANT DESIGN A pre-stressed concrete reactor vessel (PCRV) contains the total primary coolant system includin e reactorth g , steam generators, and helium circulator se reacto (FigurTh graphits . i r1) e e moderated and reflected with an active core consisting of 1482 hexagonal

61 CONTROL TOO DHVE

TOP MEAD PENETRATIONS

ORIFICE VALVES

TOP KEY REFLECTOR ELEMENTS REFLECTOE SO R

COK

COUE BARREY LKE

COR— E BARREL

STEAM GENERATOR MODULES 02!

CIRCULATOR DFFUSERS«)

HELIUM VBLVE

C«CULATO»S («)

LOWER FLOOR

FLEXIBLE COLUMNS

BOTTOM HCAO PENETRATIONS

FIGUR - FOR, VRAI1 EST T N REACTOR ARRANGEMENT fuel elements loaded with triso-coated uranium and thorium particles cast into cylindrical rods (Figure 2). Helium, at 700 psia (100% f rateo d power) s dischargei , d from four helium circulators and passes down through the core picking up heat froe fissioth m n process. This heas thei t n distributed equally through twelve steam generator modules. With the exception of the unique once-through steam generators, e secondarth e yplanth s i simila sidf to eo t conventionar l fossil-fueled units (Fig. 3). Main steam leaves the steam generator modules and enters the throttle of the high-pressure turbine with 2400 psi, 1000°F conditions. Cold reheat steam then leaves the high-pressure turbine and provides the driving force for the four helium circulators, with subsequent reheat to 1000°F in the upper (reheater) section of each steam generator module. This reheated steam is then directed back to the intermediate- and low-pressure turbine d finalle condenseran sth o t y . The conventional condensate system features a full-flow démineraiizer, three low-pressure heaters d deaeratoran , . Three boiler feed pumps (two turbine-driven and one motor-driven) direct e feedwateth r throug o high-pressurtw h e heater e sstea th bac mo t k generator modules, thus completin e cycle th e rategTh .d steam conditions help provide for a full-load plant thermal efficiency of approximatel % (Ref. 39 y l) .

62 FISSILE (U-235)

FERTILE (Th-232)

FUEL PARTICLES FUED LRO FUEL ELEMENT

FIGUR - FOR . 2 EVRA ST T FUEIN L COMPONENTS

REFUELING PENETRATIONS 1000°F

PCRV

STEAM GENERATORS

9EAERATINGJ=>- BEATES T \

HELIUM CIRCULATORS •cxy •oocJ U-ö-o— FEED HEATERW F S PUMPS FW HEATERS

FIGURE 3 - FORT ST. VRAIN SYSTEM

INITIAL TESTING Many of the primary system components of FSV are prototypical in nature and this plant has undergone an extensive and elaborate testing program including: 0 pre-operational evaluation of individual systems prior to initial reactor fueling in 1974 0 low power physics tests to qualify the design and operability of the reactor and some primary coolant components, at low power leveld an s

63 n initiaa 0 l rise-to-power progra o evaluatt m d verifan e y all plant systems through a step-by-step power escalation program. The unique design of the components within the primary coolant system, especially wite heliuth h m circulators, resulte n somi d e dela n i initiay l plant start-up. Major example f o equipmens t deficiencies involve o situationtw s s concernin e heliuth g m circulator pel ton wheels. Each circulator uses a pel ton wheel as its back-up powe re eventh driv n ti e main stea s unavailablei m t becamI . e e evideninitiath n i tl testing program that corrective measures were needed to prevent both cavitation and stress cracking of these wheels e correctivTh . e measur o t prevene t cavitation involved the design and installation of a pelton cavity pressurization system. Preventio f o stresn s cracking involved replacemen e casth t f o t Inconel wheels with wrought Inconel material. Reactor performance was closely evaluated throughout the low powe d rise-to-powean r r testing programs. Wite exceptioth h f o n a phenomena known as "core fluctuations", reactor performance was in close agreement with design requirements. Core fluctuations became prevalent as power level was increased towards 70% with a corresponding increas n reactoi e r differential pressure. This phenomenon was to be the major cause of testing and the design and installatio f o specian l instrumentatio d in-cornan bot- ex eh which required 2% years. The core fluctuation phenomenon was caused by small movements in fuel and reflector elements (moving as a column) caused by combined thermal-hydraulic effects. These small column movements resulte n redistributioi d e corth ef o coolantn d an , corresponding significant change n individuai s l fuel region outlet temperatures and steam generator module inlet temperatures. Consider n Figuri , , thae 1 f core separato th s madt i ep u e e refuelling regions wher e heliuth e m flow through each region ca n be individually regulated (orificed). This, coupled with a design where the elements are fixed in place at the bottom of the core e topth t ,a t witbuno th clearances between element r thermafo s l expansion and refuelling, set up the probability of lateral internal core movement. Non-uniform temperature d heliuan s m flows primarily between refuelling regions caused localized thermal bowing of the fuel columns which resulted in redistribution of the coolant flow. This redistribution would then caus a efurthe r localized change in the temperature profile and, over a time span (averaging about 0 min)1 , column bowing woul w patterdne occua nn i rthu s resulting in fluctuations of region outlet and steam generator module temperatures. These core fluctuations were corrected in late 1979 when constraining devices were installed on the upper elements between each refuelling region of the core. Subsequent testing to 100% power has shown that the FSV core experiences a small movement resulting in a slight redistribution f somo e individual fuel region coolant outlet temperature t higa s h power levels. However, with installatio e constraininth f o n g devices, this redistributio t cyclino n s naturi i c n d doet an eno s represen plana t t safety issue (Ref. .2) The rise-to-power testing program continued after installation e constraininth f o g devicen Novembeo d an , s19816 r V obtaine,FS d

64 100% power operation. Plant performanc t 100a e % powe s generallwa r y trouble thermaa fred an e l efficienc f 38.5o y s achieved%wa . The majority of the problems at FSV have been associated with mechanical hardware such as valves and pumps. System complexity, improper valve selection, inadequate pump and compressor capacity, r o electrical malfunctions have e beeprincipath n l reasonr fo s unfavorable plant performance; whereas, wite exceptioth e h th f o n core fluctuations e nucleath , r s auxiliariereactoit d an r s have performed quite well (Ref.3). PLANT OPERATION e first-of-a-kinTh d natur f mano es resulteyha systemV FS d t a s in reduced plant reliability. Although the helium circulators themselves have performed well and have operated at their full design values e complexitth , e heliuth f o my circulator auxiliaries and corresponding operational problems have been major concerns. Unfortunately ,a larg e numbe f plano r t transients have been initiated y b improper response/interaction betwee e heliuth n m circulators and their auxiliaries. These transients, as well as operator errors, have resulte n sizabli d e quantitie f o circulatos r bearing water entering the primary coolant helium or contaminated primary coolant helium entering the auxiliary components, with possible release into the reactor building (Ref. 4). Cleaning of moisture from the primary coolant has been a principal contributor towards low plant availability. Figure 4 is a graphical representation of plant power output froe beginninth m f 1980o g . Plant operatio o 198t n s generall4wa y good with shutdowno refuetw a r majol fo ingd s ran s modification to the circulator auxiliaries which essentially provided isolation so that a circulator upset in one loop would not affect the circulator e oppositth f o s e loop n mid-198I . n upsea 4 t occurred e plane thirty-seveoth nth t f whero x si e n contro d pairro l s required

P I I I 1 1 I I 1 1 I I I I M 1 1 I I I I I I I 1 1 1 1 I I 1 1 1 1 I I I I I I 1 1 I I I I I I I I I I I I I I 1 1 I I I I I I 1 1 I I I I I I 1 1 I I I I I I 1 1 I I I I I I I I I 1 ! I ! I I I

1980 1981 1962 1983 1364 1985 FIGURE 4 - FORT ST. VRAIN GENERATION 1980-1986

65 manual operator action for insertion after receipt of a reactor scram command a result s A l thirty-seve.al , n contro d drivro l e mecanisms were subsequently extensively refurbishe d broughan d t a "near-new o t " condition e controTh d .driv ro l e refurbishment program was completed in July, 1985. However, extensive discussions had bee progresn i n s since January, 1985, witr Nucleaou h r Regulatory Commission (NRC) pertainino meet e abilitV t th FS regulator o f t o gy y requirements (10CFR50.49) associated with Environmental Qualification f electricao l equipment (EQ). e latIth ne 1970's ,a sizabl e effor s undertakewa t o qualift n y plant electrical equipmen e harsth o ht tenvironmen t involvina g high-energy steam line break. This effor s documentewa t d an d e 1979-198th n i C 0 NR timsubmitte e eth frameo t d A subsequen. t review by the NRC of our EQ Program in late 1984 led to three areas f concerno . These included equipment aging, equipment operability time d operatoan , r response timen conjunctioI . n with operator response time e testinth , g progra o t qualifm y equipmene th n i t late 1970' s basewa s d upoe operator'th n s abilit o t isolaty e a high energy steam line break within four minutes. The subsequent temperature profile in the reactor and turbine buildings resulting from four minutes of blowing steam was utilized in the equipment testin s subsequentlha g C programNR e Th y . determined that credit could not be given for proper operator response in less than ten minutes. The good steam conditions afforded by FSVs design made the temperature profil a ten-minut r fo e e isolation time prohibitive in qualifyin r equipmenou ga progra d an o tdesigt m d instalan n l a very fast steam line break detectio d isolatioan n n systes wa m undertaken. Installatio f thio n s system allowo takt s eu s credit for previously tested equipment as environment temperatures now resulting fro a mstea m line brea e substantiallar k y lower. The plant is currently shut down to complete the Environmental Qualification effort. Start-up is anticipated for January, 1987. RADIOLOGICAL PERFORMANCE The radiological consequences of operating FSV have been exceptionally low. The personnel exposure rates and discharges e environmenth o t t have been consistently order f magnitudo s e below those of the average light water reactor (LWR) power plant in the United States. The major reason for this performance is inherent e cord e primarth desigan th ie nf o n y coolant system. Release of contamination into the reactor building environment is minimized e uniquth ) e (1 bygraphit e core structure, with ceramic-coated fuel particles, (2) the use of inert primary coolant gas, helium, e PCRth V ) anwits doubl(3 dit h e sealed pressurized (with purified helium) penetrations. The equilibrium radioactivity circulating in the primary coolant was conservatively calculated to be 30,900 Ci (design level) with an expected value of 2630 Ci. Actual measurements at 100% power show 439 Ci with the third cycle core (Ref. 5). Figur 5 providee a s comparisospersonne V FS f lo n exposure history with that of the U.S. light water reactors. The radiological exposures for each of the years 1980 through 1983 was less than 1 man-rem grand total for all personnel badged at FSV. Total

66 1900 1961 198Z 1363 1964 1385

TERR FIGUR AVERAG- 5 E E PERSO M EXPOSURRE N BHRR FO E ,V FS PWR D AN , exposure levels increased to 35 man-rem in 1985 due almost entirely e controth d refurbishmeno ro t l t program o farS . n ,198i 6 total personnel exposure has been less than 1 man-rem. A total of 150 m3 of low-level solid waste has been accumulated at FSV in the past 12 years. This includes waste generated as a result of (1) initial fuel loading, (2) three refuel ings, each with approximately one-sixte reactoth f o rh core replacedd an , (3) numerous plant modifications and maintenance operations, including major work performe n componento d s withi e primarth n y coolant system. With the exception of tritium, noble gas airborne and liquid effluent releases V frohavFS me been greater than ordea n f o r magnitude belo e e averagU.Sth th w .f o nucleae r power industry. Higher than anticipated amounts of tritium have been experienced V primarilFS t a y becaus f e upsetheliuo e th n i sm circulator auxiliary system, which have allowed small amount f bearino s g wate o entet r r and circulate in the primary coolant system. The tritiated moisture is removed by the purification system and is discharged from the plant as liquid waste (Ref. 6). FUTURE Although the last two years have been difficult for FSV because of the control rod drive and EQ programs, Public Service Company of Colorado (PSC) has been committed to making FSV a viable electricity-generating plant. To help accomplish this, the FSV Improvement Committe s formewa e o t investigatd e possible plant modifications which would enhance plant availability. Moisture ingress eminating from the high pressure bearing water within the existing helium circulator s e beeprimarha th sn w y lo caus r fo e plant availability and is therefore the prime candidate for review by this Committee. Possible actions currently being considered e th Committeby e include investigatio e th followin f o n g possibilities:

67 0 Replacemen e existinth f o tg circulators with machines that utilize magnetic thrust bearings while continuin o incorporatt g e the existing pelton wheel and single stage steam turbine drives. f magnetio e us ce Th thrust bearings would allo e eliminatioth w n f higo h pressure bearing water; however, verw pressurlo y e water woul e use b s d a bot e dlubricatin th h g e mediuth n o m journal bearings as well as a sealant to prevent the primary coolant from flowing down the circulator shaft. 0 Evaluate the possibility of extensively modifying the existing circulator auxiliaries in the following manner: - complete separation of bearing water and buffer helium systems for each of the four circulators - eliminatio f back-uo n d emergencan p y bearing water sources by incorporating individual bearing water pumps for each circulator powered by Class 1 electrical systems. 0 We also may elect to evaluate the possibility of installing hydrostatic seals on the existing circulator shafts which would tend to seal the bearing cartridge from the primary system upon a pressure differential reversal. If the decision by the Committee is to proceed on these evaluation efforts, they will continue ovee nex 8 th monthsr1 t . Final selectio f circulatoo n r improvement se anticipate b woul t no d d until these reviews have been completed (Ref. 7). Poor operation of FSV throughout the past two years has resulted in PSC's recent financial evaluatio f thio n s Plant with respect to overall Company economics. This evaluation resulted in the following actions: 0 A write down of PSC's asset in its Fort St. Vrain Plant involving approximately two thirds of the current capital cost associated with the plant. 0 Removal of Fort St. Vrai n from the rate base. However, a partially escalatable rate of 4.8

68 CONCLUSIONS Considerable effor s beeha tn expended ove e paso yearth rtw t s to bring the plant to a high degree of readiness. This includes enhancement in the areas of management, training, increased technical staffing, preventive maintenance d proceduran , e controlsC PS . is confident that successful continuous plant operation wile b l achieved beginnin earln i g y 1987.

REFERENCES

[1] Brey H. L. and 01 son H. G. Fort St. Vrain Experience. Nuclear Energy Apr., 2 ,. 1983,No 117, ,22 . [2] Ibid, 117-118. , d Oison6. an . . , ] H BreyL , KantoMoore E. [3 A. . H . ,. M R r, HT6R Experience, Programs d Futuran , e Applications, Nuclear Engineerin Designd an g 2 (1982,7 ) 167. [4] Ibid, 167. [5] Brey, H. L. and Graul, W. A., "Operation of the Fort St. Vrain High-Temperature Gas-Cooled Reactor Plant," Proceedingf o s the American Power Conference, Chicago, Illinois, April 26-28, 1982, 5. [6] Ibid, 6. [7] Brey, H. L., "Fort St. Vrain Update", presented at the Eighth International Conferenc e HT6R th n Diego Sa n ,o e , California, Septembe 1986, 15 r .

69 AVR EXPERIENCE

E. ZIERMANN Arbeitsgemeinschaft Versuchsreaktor (AVR) GmbH, Jülich, Federal Republi f Germanco y

Abstract

e papeTh r contain a descriptios e Arbeitsgemeinschafth f o n t Versuchsreaktor (AVR) a smal, l pebbl d HTGbe e R locate Jülichn i d , Federal Republi f Germanyo c . Beside a descriptios f plano n t designe th , experience made with main components e operatinth , g performancd an e operational result e presentear s d together wit n outlooa h o possiblt k e future plant improvements.

. 1 Introduction The Arbeitsgemeinschaft Versuchsreaktor (AVR) GmbH is an associatio 5 electri1 f o n c utilities establishe n 1959i d . The aim of this association was to obtain scientific, tech- nologica d economian l c information abou e reactoth t - sy r stem from the construction and operation of an experimen- tial nuclear power plant wit a high-temperaturh e reactor. In August 1959, the order was placed with the company partnership BBC/Krupp for the construction of the experimental power station e plan. th Wor tn o kbega n i n 1961 n AugusI . t 1966 e reacto,th r became first critical, ann Decembe o , d1967 th 7 1 ,r electricit s suppliewa y o t d e publith c suppl e firsyth grir t fo dtime n FebruarI . y 1974 e meas outleth ,ga n t temperature, whic d initiallha h y . s increase°C wa 0 , 95 °C o t d0 85 t a beet se n The f initiaexperimentao m ai l l operatio o demont s -wa n strate that a helium-cooled pebble-bed reactor can be constructed and operated safely. In addition to continuous componen te collectioth testin d an f o goperatinn g experience and results, the plant was also used in the following period for the implementation of experiments. In particular, these included the testing of various types of fuel elements, investigations of fission product behaviour e circuiith n s wela ts experiment a l e chemistrth n o s f o y coolant gas impurities.

71 reactoR AV FIG. e rVie1 th building. f wo .

Structure of the plant

Th e reacto th cor f o er consist pebble-bea f so d containing 100,000 fuel sphere sgraphita witn i a diametehm c e 6 f o r reflector pot. The steam generator is located above this s shieldei cord an e d against radiation froa e cor y th mb e m thic m p reflecto 0 to k 50 r mad f graphito o addie tw d - an e tional 500 mm thick layers of carbon brick. The coolant is helium which is pressurized at 10 bar and circulated by o blowertw s e locateloweth rn i d sectio e reactoth f o nr vessel. The helium is heated in the core from 270 °C to d thean n C ° flow 0 s95 throug e steath h m generator, where it transfer e water-steae energth th s o t y m circuite Th . generated steam has a pressure of 73 bar and a temperature e coreTh ,. °C stea 5 mo50 fgenerato d bloweran r e ar s surrounde o concentricalltw y b d y arranged reactor vessels. The interspace between the vessels is filled with pure

72 FIG. 2. Sectional view of the reactor building.

helium at a pressure slightly above the coolant gas pressure n initiaA . l biological shiel s locatei d d between the cylindrical sections of the reactor vessel. This double vessel system is again surrounded by a safety containment and a 1.5 m thick concrete tower. The reflector, made from graphite blocks s fouha , r bored bulges, which project inte corth oe space e shutdowOn . n ros fittei d d into eac f theso h e bores. These shutdown rods have the task of compensating for the differenc reactivitn ei cold an yd t betweecriticaho e th n l condition o keet e reacto th d p ,an r subcritica e colth dn i l

73 condition by at least 0.5 %. During power operation, the fuel elements are continuously circulated and measured regarding their burn-up. Spent fuel elements are removed and replace y fresb d h fuel.

In the purification plant, the helium is purified of dust as well as active and inactive impurities.

. 3 Experience witmaie th hn components

3.1 Fuel

The fuel elements have an external diameter of 6 cm. In an interna m diameterm l0 5 zon e fuef o eth s ,locate i l n i d the form of 30,000 coated particles. These particles have a kerne f severao l l tenth a millimetr f o s n diametei e r consisting of mixed crystals of uranium and thorium carbide, or uranium and thorium oxide. The kernels are surrounded by several layer f pyrolyticallo s y deposited carbon innee Th . r layer has a relatively slight density and is compressible.

FIG. 3. AVR — Fuel element.

74 By compressing this layer, storag e efissio th spac r nfo e gase s producedi s e externaTh . le pressure th laye s i r - bearing encapsulatio e particleth f o n nA additiona . l diffusion barrier layer is partially provided between these layers. Wit hw exceptions fe onl a y e fueth ,l elements contain l g uranium-235 and different amounts of thorium and uranium-238.

Overall, fuel element f differeno s t composition d paran s - ticle sizes were teste n largi d e numbers under operating conditions up to high burn-ups. This showed that fuel elements with excellent fission product retention capacity can be produced. The average burn-up is noticeably above r tonn100,00pe ed MW 0heav y metal, reachin a maximug f o m slightly more than 180,000 MWd per tonne heavy metal. Our e verw coolanth lo y s t activita ga t , Ci f abouo y 5 2 t present time, is a consequence of these very dense fuel elements.

2 Reflecto3. r

The inner sid p reflectoe to reflecto e th rd an consisr f o t needle coke graphite ARS/AMT manufactured by Sigri-Elektro- graphit. The side reflector consists of extruded and the

reflectorp To — .R FIGAV . .4

75 top reflector of rammed blocks. Since the upper side re- p reflectorto e flectoth d e ,fue an th rnex l o t elementst , are exposed to high neutron radiation, there were fears o t neutron-induce e thadu t d deformation stressind an g cracks could have occurred in the reflector. It was also feared that there would be corrosive material erosion due to coolan s impuritiesga t . These concerna visua o t ld le s inspection of the top reflector and the upper side reflecto n 1984i r . This p reflektoinspectioto e th rf o n and the visible sections of the upper side reflector showed thae reflecton th perfeci t s wa tr condition. Cracks t werfoundno ed corrosioan , t no n s weawa r detectable.

3.3 Fuelling_system

The fuelling system permits insertion of the spheres against a helium pressure of 10 bar, the removal as well e circulatioth s spheree a th f o ns during power operation. The transfer device e formear sy b bald l valved an s retaining latches e circulatioTh . n route starts wite th h sphere discharge tube ,n interna a whic s ha h l diametef o r o avoit m m d bridg0 50 e formation. Thi s closei a s y b d slotted disc, the so-called diminisher. By turning this disc, spheres fal n groupi l s inte storagth o e container of the singulizer. This singulizer is a scoop wheel, which raises the spheres out of the container and supplies them to the scrap separator. This is where defective sphere d fragmentan s e separatedar s t presentA . , the fracture rate is about one defective sphere per 30,000 circulated spheres e intacTh . t sphere e passear s d to the conveyer via a dosing wheel. At this conveying point, the type and burn-up of the spheres is determined. On the basis of these results and input control data, a computer then decides in which of the five sectors the spheres are to be resupplied to the core, or whether they mus e removeb t d from circulation. Thise th syste s ha m considerable advantage that the burn-up can be continuously balanced by the addition of fresh fissile material. In addition, there is no necessity to shut down aa specifit c timr refuellingfo e .

76 This system was developed from scratch for our power station and incorporates errors typical of a pioneering system of this type. Thus the different friction

a) Top reflector 1965.

b) Top reflector 1984. (VIDEO INSPECTION)

FIG. 5.

77 FUEL HANDLING SYSTEM

> REDUCEff R

SINGULIZE© R

iSPHERE VALVE

i FRAGMENT VESSEL

DOSING WHEEL

CONVEYOR

FUEL ELEMENT VESSEL

) RIN(? G ROOM

FIG. 6. Fuel handling system.

conditions unde r r wert takeheliuai no e d nan m into adequate account, and the design of the individual components was not optimized. Despite these understandable deficiencies, the system performans satisfactorily. Up to now over 2.2 million spheres have been circulated l defectAl . s were rec- n tifieacceptabla n i d e time witw radiatiolo h n exposure for the repair personnel. In part, these repairs were

78 carried out without interrupting power operation. Only less e non-av/ailabilitth f o tha% 3 n y tims causewa e y thib d s fuelling system.

3.4 Shutdown rods The reacto s fouha rr shutdown rods, whic e insertear h d into the reflector core from below. They are arranged symme- tricall circla n i y e wit diametea hwhos, m 2 e f centro r e point coincides with the vertical axis of the reactor core. The rods move in bores of the four graphite noses. Each shutdow s connectei d a heaviero n o t d r counter-roa vi d a rack- and -pinion, so that when the coupling between the pinion and drive is opened, the absorber section is pushed inte corth oe from belo y gravityb w e shutdowTh . n rods a s such consist of concentric tubes with boron carbide filled in between them.

SHUT-DOWN ROD

W CLUTCJA H

PINION

COUNTER ROD

SHUT-DOWN ROD

FIG. 7. Sectional view of the AVR reactor building.

79 The d dromotioan p ne shutdowth time f o s n rode ar s measured. Up to now no changes in the motion and drop characteristics of the rods have been noted. The rods themsel- s havve e operated maintenance d trouble-frean - e since th e commissioning of the system. The gearing of the racks and pinion s checkewa s n 1980i d . Here again o defectn , s were e e rotarregionotedth th f n o yI n. penetrations e havw , e replaced the metal bellows, as the theoretically expected service life had been reached. As a precaution, all bearinge drivth en i ssectio n were also replacee th t a d same time.

3.5 §team_generator

The steam generator consists of four parallel Benson (once through) systems. The temperature is controlled by water injection outside the reactor vessel. In front of this injection, each system consist f fivso e parallel tubings.

— Stea R mAV . FIGgenerator8 . .

80 The superheating section behind the injection consists of ten parallel tubings per system. Thus the steam generator consist 0 separata tota6 f f o so l e tube loops e proporTh . - tion of superheater tube surface is 5 % of the total heating surface o avoiT . d power losses through possible failuref so tube loops, the steam generator was designed with a standby heating surface of 10 %. The tubes of the steam generator are made from ferritic steels, as are normally used in conventional boiler construction. After 72,000 hour trouble-fref so e operation ,leaa k occured in one final superheater tube. Even though the cross-section 2 hole s finalloth fewa y onl7 tonne2 y, wate f abousm o m r3 t entered into the primary system, flowed through the core and collected in the lower vessel section. The inflow of such a large amount of water was possible, because firstl e chemicath y l compositio leakage th f o ne water resulted in wrong conclusions as to the source of the leakage, and secondly no measuring instruments were available which could have indicate e volumth d f penetratino e g water. Whilst the location and closure of the defective steam generator tube took place without particular difficulty, the removal of the penetrated water from the primary circuit and the drying of the system required a considerable amount of time. The difficulties during the removal of the water were mainly e laccause drainagf th o k y b d e possibilities. Thusa residue of 2 m water could only be removed by evacuation.

Th ee lea causs th stili k f o el unclea t presenta r , since we have not removed the steam generator due to the expense this would incur. However, failure due to creep is excluded, since the steam generator was so conservatively designed that up to now only a slight portion of the theoretical life expectanc s beeha yn utilized. Since completion e repaith f ro word recommissioninan k e plantth o f n o ,g furthe re servic th damag w s occurredno eha e lifo t ep U . bees ha n slightl excesn i y 111,00f so 0 operating hours.

81 CORE

COOLING GAS BLOWERS

SPHERE DISCHARGE SHUT-DOWN TUBE ROD

REDUCER

CONVEYOR

Wate— R rprimar e AY leveFIGth . n 9 .i l y circuit.

3.6 Blowers

o coolanThtw e s blowere lowega tth n ri ssectio e th f o n reactor vessel each consists of an asynchronous motor, mounte n oil-lubricatei d d plain bearings n impellea d an , r mounte n overhuna n i d g positio e mototh rn o nshaft . Laby- rinth a seas syste d ga l an s m preven l frooi tm enterine th g coolan s circuiga t d coolanan t s froga tm enterin- mo e th g tor casing e spee Th .s adjuste i d a frequenc a vi d y changer. The blower e designear s r maintenance-frefo d e operation. Wite exceptio th he bearin on f o ng exchang n 1978i e , which required the removal of one blower, these machines have been operating maintenance-free since 1966, even though after the steam generator defect both machines were submerged in water above their axles.

82 t steaCu m— generatoR AV . FIG10 r. pipes.

FIG. 11. AVR — Welding of steam generator pipe to steam collector.

83 The removal of this one blower was necessary because, after maintenance work on the lubrication system, cooling water had penetrated into the bearing oil system through a leaking screw union.

Removal of the blower and replacement of the bearing did not caus y majoan e r problems. Durins e workwa th gt i , noted that the blower was in very good condition, and that it showed no damage due to ageing.

BARRIE 1 INPU S GA TR

BARRIE 1 OUTPU S GA R T

1 SHAFT 4 LABYRINTH GLAND 2 ELECTROMOTOR 5 BEARING 3 ÎMPELLER

FIG. 12. AVR — Cooling gas circulator.

3.7 ^essels_and_supporting_structures

The s subinnei t -I r . reactoMo + 5 r n vesseM 9 1 mads i lf o e jecte o virtuallt d o stresn y s during operatione th s a , interspace between the vessels filled with pure helium, is subjec o almost t e samth te pressur e coolann i th s s a ega t the inner vesse s sucha l . Slight differential pressures only occur durin e start-uth g d shutdowan p n processes, whice ar h far below the design pressure. Up to now, the inner vessel has only been actually stressed twice during pressure testing.

84 Dismountin— R AV blowere FIG . th f 13 g.o .

FIG . Damage14 . d radial bearing.

85 The outer reactor vesse s madi lf fine-graio e n 5 stee4 B MS l wit a thermah e truTh le . yiel°C d 0 0 N/mm20 poin25 t a f 2o t load is below 200 °C at 127 N/mm2, so that no limitation applies here either.

The supportin ge graphit th structur e cord d th an e an e r fo e carbon s designei bric t I k . d3 internal o M 5 1 s madi sf o e for a temperature of 450 °C. The actually measured tempera- o dimensionat e Du . °C l 0 stabilit27 tur s i e y requirements, the limit shoul 0 t deflect 47 no dloa structure f o dth t y b e more than 1.5 mm. This requirement leads to exceptionally w streslo s loads e supportinTh . g structur s protectei e d against radiatio e nspher th froe cord th eman e discharge m thicc 5 1 k a steetub y b el base plat d shieldean e a y b d

REACTOR VESSELS

SUPPORTING STRUCTURE

FIG. Sectiona15 . reactoR AV le rvie th building f wo .

86 boron carbide shiel o s e thadneutroth t n doss i e significantl m yc evebelo n0 1 wafte 9 year11 r8 f operaso —2- tion. This means that this component is also operated at well below its design values, so that a service life limitatio t expectedno s i n . purificatios Ga 8 3. n system

Through the addition of fresh fuel elements, through minor leaks in cooling equipment and experiments, as well as during maintenance and repair work on primary circuit components, air and moisture can enter into the primary system. To limit graphite corrosion e leveth ,f impuritie o l s i s kept low by continuous purification of the primary circuit. The suspended matter is retained in a gas prepurification system, which consist o paralletw f o s l grave d filterbe l s and coolers. The throughput of approx. 800 m /h is produced by the delivery pressure of the coolant gas blowers.

STORAGE VESSEL BARRIERGAS2

ADSORBER H20

Purificatio— R AV . FIGn16 . plant.

From the helium prepurified in this manner, a bypass flow of 50 m /h is extracted and passed through two purification systems arrange n tandemi d e system e firsth Th .f o ts consist f silicageso l drier activated san d charcoal adsorbers, which are kept at low temperatures using liquid nitrogen.

87 This sectio s operatei n statn i d f adsorptioo e r approxfo n . 3,000 hours, so that mainly noble fission gases are retained here. In a catalytic oxidation stage, hydrogen and CO is oxidized e subsequenTh . s purificatioga t n stage also consist f liquio s d nitrogen-cooled adsorbers, whice ar h operated in state of adsorption for a mean time of approx. 10 days. In addition to the oxidation products, mainly nitrogen is retained here. Both gas purification stages are duplicate o thas d t whee stag on ns loaded i e a changeove, r to the unit connected in parallel is possible. The loaded stage is desorbed by heating, and the desorbed substance is retaine n delai d y tanks unti e vasth lt majorite th f o y short-lived noble gases have decayed. Initially, we had problems with the diaphragm compressors which pump the gas through these systems. After SCIÏIR of these compressors had been replaced by dry-running piston compressors, the systems now operate to our full satisfac- tion.

4. Operating performance

1 Powe4. r control Our reactoo contron s ha rl rodr powefo s r control teme .Th - perature of the reactor is adjusted via the fuel loading, e powean th ds adjustee blowei r th a rvi d speed, i.ea .vi the helium mass flowd thuan , s throug e energth h y extraction froe core th e powem Th . r contro s effectei l d automaticall e negativth a vi ye temperature coefficient. The steam produced by the steam generator is supplied to the turbine, which is fitted with an initial pressure regulator t thuI . s takes precisel e poweth y r s i whic t i h offered by the steam generator. The steam temperature is adjusted throug e feedwateth h r mass flowe systeTh . m operates in such a stable manner that we can practically do without the planned temperature control through water injection.

88 2 Start-u4. p procedure When startin p froe colu g th md condition o limitationtw , s mus mainle b t y observed e firs.Th e tth limitatioy b t se s i n tensile strength of the carbon brick bridge above the core. o prevent s e Stensila oth t e stresses insid e brickth e s from exceeding the permissible values, due to the temperature gradients durins ga g t heatingho e th , temperature may only be raised by max. 3 K/min. The second limitation arises from stability considerations e steafoth r m generator o avoi.T d vaporizatio wrone th gt na places in the steam generator, steam production should only start when the helium entering the steam generator has a mean temperature exceedin . Fro°C m 0 thes60 g e limiting conditions a ,start-u p procedur s beeha en devised whice w h divide into five stages e firsth tn I . stage e reacto,th s i r operated until it is critical by withdrawal of the shutdown rods.

TH THERMAL POWER SH = HOT-GAS-OUG T TEMPERATURLE T E

SPEED OF BLOWERS •» = FEET - WATER - TEMPERATURE w f

SR AVERAGE POSITION OF SHUT-DOWN RODS X = WET - STEAM MASS-FLO F FEED-WATERWO 1

-1000 o o *- X x

Start-u— R pAV diagram. FIG17 . .

89 e seconIth n d stage e powe, th s increasei r thar e fa th t o s d steam generato n jus ca e rkep b t n wate i t r operation. With this power, the temperature is then raised to 600 °C in the third stag n accordanci e e wite limitatioth h n specified above. Whetemperature th n , vaporizatios reache°C ha e 0 60 d n in the steam generator is started by increasing the power and reducin e feedwateth g r mas e sfourt th flo n i hw stagen I . the fifth stage, the power, coolant gas temperature and steam temperature are set to the required values, the turbint intpu o s i operatioe e generatoth d an ns synchro i r - nized wite mainsth h . When running-u e turbinth p o fult e l capacity, limits again arise e froturbinth m e material stressing e nee.W dhala foud fan rhour r start-upfo s , from the beginnin d withdrawae synchronizatioro th f o o gt p u l f o n the generator.

3 4. Shutdown procedures Three different procedure e availablar s r shutdownfo e : 4.3.1 Scram e maxth f .I permissible neutro ne max th flu .r o admissiblx e neutron flux change is exceeded, an automatic scram takes place. The rods drop into place, the coolant gas blowers are switched off and the generator is disconnected from the mains. After such an automatic shutdown, the rods are manually withdraw o halt n f their lengto t h s a agai o s n reduce thermal stresses. To cool down the reactor system, the feedwater approxt masa . % st 0 se flo4 . s i w

4.3.2 Insertio shutdowe th f o n n rod motoy sb re drivth s i e second procedure All other activations e froreactoth m r protection system caus n automatia e c insertio e shutdowth f o n n rods, combined with a shutdown of the coolant gas blowers and the shutdown of the turbine. Even after this shutdown, the shutdown rods are withdrawn again to half their length, and the feedwater t approxmas a . Afte % t s 0 se r4 flo. s aboui w t three hours, the blower e thear s n starte p agaiu do t n stages i ns a o s , accelerate the heat removal.-

90 - AVERAGSR E POSITIO SHUT-DOWF NO N RODS IÏIM.OWER- SPEED OF BLOWERS PTH - THERMAL POWER 50-

_ J l z1 40- z — --— X £ Z -y. •x. z 30- o SCRAM X o — \ œ 20- \ * £ \ j o: û- \ H/> \ ,1 10- SR PTH s-.^nBLOWER

0- 0123456789 50- MIN-*N I T -— - i — . — T •z, 40- IE Z \ DRIVE IN 0F Z Z ^ ^v. 30- ÏN z Z \ SHUT-DOWN RODS _^™ \ -^^^ \ ^^ X *o ^ oc 20- x ï PTH\ \OBLOWER §R^\^^ g tE Q- \ NXN \ !

X "c, — o= 20- lu H ï PTH\VOWER BLOWERS ONLY 1 K. CL. V H? 10- v^___^ n- 0123456789 —T IN MIN-*-

FIG. 18. AVR — Shut-down procedures.

4.3.thire Th 3d procedur switchins i ee th f of g coolant gas blowers Switchine coolanth s blowere f mosga tth of g t s i gentls d an e slowest shutdown process s alwayi t I .e reacto s o th uset f s i di r be shut down specifically, wite rodth hs initially remaininn i g position. Only after some time they are inserted into the core position, whereactoe th n s considerablha r y cooled down wits A . h the other shutdown procedures, the feedwater mass flow is set at t e intblowerpu th o e d operatioar an sapprox , % 0 stagen 4 .i n s after three hours.

91 4.4 Since load changes only involve slight temperature variationn i s the cord reflectoran e e limitatioth ,e rat f th chango e r fo ne only arises fromaxe th m. possible chang blowen i e r speed. We have carrieo t loatw ou dn d i reduction% 0 5 o t s% fro0 10 m minutes. When raisin e powerth g , equally rapid changee ar s basically possible. However t musi ,e noteb t d that when changin e bloweth g r speed the removed amount of energy is set. The core power is automatically e readjusteth negativ a vi e d temperature coefficient e e inertisysteTh th . n mi a cause e poweth so t r overshoot. Where the power is carelessly raised, this overshoot could resul a shutdow e reactorn i o th treachint f e o n e du , th g limitin e gmax th valu . r admissiblfo e e power.

NEUTRONFLUX RECORD DATE: 02.14.1984

15 00

14 30 I- ^ THERMAL POWER 0 100% FIG. 19. AVR — Dynamic experiment.

92 4.5 Performance during failure of the afterheat removal 5y§tem_and_the shutdown_rods____ We have demonstrate f experimeno y wa y b dt that after failurf eo e activth e heat removal during power operation, with simultane- s failurshutdowe ou th f o e n device reactoe ,th r shuts down automaticall d remainan y s subcritica r abou fo e dayl on t e .Th temperatures tha te fuearisth ln i remaie n withi e rangth n e of temperatures prevailing during operation. Calculations by computer show that with a simultaneous loss of coolant gas the max. fuel temperatures woul t reacno d h critical levels.

As described above, we have taken this event, which was originally considered to be a serious emergency condition, as one of our standard shutdown procedures.

TEMPERATURES : 1 THERMAL POWER t. REFLECTOR NOSE TOP 2 SPEED OF BLOWERS 5 REFLECTOR NOSE MIDDLE 3 AVERAGE POSITION 6 SIDE REFLECTOR INSIDE OF SHUT-DOWN RODS 7 REFLECTOR BOTTOM

800i

700' —— 'SECONDARY CIRCUIT OUF ACTIOO T N

600'

50QJ

400-

300'

200

PAVERAGI CA 300kV

100- _ .1 _ 0 2 26 •T IN IH- Simulate— R AV . dRIG failur20 . f shut-doweo n equipment and interrupted decay heat removal for that time.

93 5. Operational results

5.1 Sinc firse eth t suppl electricitf o y Decemben o y r 17th, 1967 until today, 1.5 thousand million kW hours of electric power have been produced. Thi s equivaleni s o 4,10t t 0 full-load days. The mean time utilization is on average approx. 68 %. We achieve bese th dt figur numbee 197n i eTh 6f . o r% wit 2 9 h shutdowns due to incidents was noticeably reduced during the first few years, and since then has fluctuated at figures of less than 5. Sinc e havw e e focuse r attentioou d e demon th t onl n no n-o y stratio f gooo n d operating result d componenan s t testing, but for some time increasingly on the implementation of experiments, which we will still intensify in future, the time utilization in the coming years will be slightly reduced.

tosrf For elo

W68t 69 I 70 I 71 \ 72 \ 73 1 74 I 75 I 75 77 79 I 80 I 81-1 82 \ 83 I 54 I 55 I §6

FIG. 21. Availability of the AVR power plant over the years 1968-85.

94 t in a- . Shut-downFIG 22 R .powe AV e rth planf so t ove yeare rth s 1968-85.

2 R5diation_exposure_to_the_personne5. l The collective e personneannuath l lal doslr employefo e d varies between 0.5 and 0.6 Sv (50 to 60 man x rem). All in decreasina s ha alt i l g tendency.1 9 yearDurin1 f e operatioo sth g permissible th n e individual doses have so far never been exceeded.

5.3 is radioactive substances The emission of radioactive substances into the atmosphere is slight. It ranges between 15 and 30 Ci/a for noble gases, about 50 Ci/a tritium and 1 to 2 Ci C-14. The emissio f aerosolo n s withii s e |j.Cth ni range, whilst iodin s virtualli e t presene exhausno yth n i t t air.

95 5 8 4 8 3 8 2 8 1 8 0 8 9 7 8 7 7 7 6 7 5 7 4 7 3 7 2 7 1 7 0 7 9 6 8 6 —— t in a ——i»-

83'84'82 8 1 8 0 58 9 7 8 7 7 7 73'72 6 7 7 1 5 7 47 0 7 9 6 8 6 —— t in a —»•

FIG. Dose.23 personneo st R poweAV n lri plant over the years 1968-85.

Possible improvements as seen by the operator e froNaturallyr planse ou mn t ca tha e w ,t improvemento t s both the design as well as the individual components are possible. Thus the division of the biological shield into two section a definit s i s e disadvantage e consequencTh . e of this i thas t numerous facilities which require maintenance, such as measuring equipment, valves, compressors, etc., are located in areas exposed to increased radiation. The relatively poor top shielding of the reactor buildin s alsi g disadvantageoa o reflectiot e Du . n radiatioe o th fatmosphere th n o n e (sky-shine effect) this lead o areat s f slightlo s y increased radiation fieldt a s the power station site.

96 The occasional difficulties with the fuelling system have already been mentioned. Similarly e initiall,w d diffiha y - culties wite diaphragth h m compressors, solenoid valves, the frequency coolane changeth blowers r ga tfo r , etc. However, we are also aware that our experience has been fully taken into e accoundesigth f subsequenn o ni t t plants o thas , t these item e AVR-specificar sy b e ar d an , no means characteristic of an HTR.

. 7 Summary Our AVR experimental nuclear power station is the first plant wit a helium-cooledh , high-temperature reactor using spherical fuel elements. Despit e resultanth e t imperfec- tion difficultiesd san plane s buil,th relativela wa tn i t y short time, and has now been successfully operated for over 19 years. It has been found that the reactor has excellent safety characteristics, and that the radiation exposure for the personnel as well as for the environment is very low. Operatio e plan th s unproblemati i tf o n d doet makan cno s e any increased e demandoperatinth n o s g personnell Al . disturbances that have arisen were mainly rectified by our n operatinow g crew. The conditio e maith n f o componentn s doe t indicatsno y an e o changeageingt e r operatindu Ou s. g experience th d an e pre-exposure of the plant do not suggest any limitations to the further use of the plant.

97 THTR OPERATION — THE FIRST YEAR

D. SCHWARZ Vereinigte Elektrizitätswerke Westfalen AG, Dortmund, Federal Republi f Germanco y

Abstract

e papeTh r contain descriptioa s performance th f THTR-30e o n th f eo 0 since its first criticality in 1985. The THTR-300 is a pebble bed HTGR of 300 MW(e) located in Schmehausen/Uentrop in the Federal Republic of Germany. The main steps and events during the start-up and commissioning phase are presented and discussed in detail.

Main features and overall valuation

One year ago, Sept. 6, 1985 the THTR 300 had reached first criticality.

One year late y Sept(b r, 1986) .6 reactoe ,th bees ha rn critical 3 hours15 5 . r Firsfo t curren n Nov o s sen , 198t a .twa 16 tou t 5a % hav 0 powee6 bee; % r n0 leve 4 reache f o l Marcn o d, 1986 19 h , 80 % on April 10, 1986. The power plant has produced a total of 0 millio42 withih 1 hourKW n07 Septy 3 b ns , 19866 . .

In the meantime, on Sept. 23, 1986 THTR has achieved the 100 % power s beelevelha nt I operate. t thaa d t level till Oct , 1986.3 . Thet i n has been shut down for a 6 weeks-revision, after having produced a total of 540 million KWh.

Comparing THTR commissioning with tha f otheo t r nuclea cond an r - ventional power plants we can state a very satisfactory lack of problemg bi y t commissioninan Ye . g took much more time than with plants that had big problems. The main reason for this seeming con- tradictio e facth ts i nthaprototypea s ti THT0 30 R .- Thiex s i s pressed in

the great number of experimental transients required, by far exceeding one hundred

99 the problem f stayino s g between very narrow design marginf o s temperature, pressure, flow etc. during those transients, leadin o changet g proceduren i s d repetitionan s e experimenth f o s - tal program

small, however annoying problems with components, controld ,an operator errord ,an

special carefulnes e authoritth f experto f o sd an y s actinn o g behalf of the authority, leading to time consuming persistence on formal procedures, on the experimental program, and on technical margins. Practically every activity needs expert opinion and the authority's consent, for major steps in the commissioning program we have to get a release, and for changes that have more than zero relevance with respec o safett w licensee neene yw a d . Needless to say that much time is needed for preparing and getting decisions and for the paperwork before and after every activity.

In addition, Chernobyl cost us a total of 61 days, 22 due to an interruption of the experimental program (operation at steady power) a politicall to due y39 motivated shutdown.

In view of these circumstances, the progress in commissioning THTR as it has been achieved up to now can be valued as very satisfactory.

THTR operation from Sept. 198 Septo t 5 . 1986

Let us now analyse the power raising history of THTR in some more detail . Thers give2 a , fign d i enan hav.1 e bee 7 event3 n 5 shut(3 s - downs and 2 load reductions) since first criticality (the 37th at the beginning of Sept. 1986): 13 shutdowns and 1 load reduction can be attributed to instrumentation and control (hard- and software); 10 ca- s havse o operatoet beee du n r error, often involving control problems; in another 10 cases the plant had to be shut down for mechanical reasons in the conventional part of the plant, mostly leaks. 2 shutdowns resulted from planned experiments, 1 load reduction had an external cause.

100 h î rTbermaP l power :Bot gas temperature P# :Sectrfc power [TM« «fein «team t«*f»«»i»«»

FIG . Performanc1 . e diagram during commissioning (THTR 300) (Phas Februar— ) eLI y 1986

:Thermai power :Hot sas temperature

:Bectric power :Main steam temperature

FIG . Performanc2 . e diagram during commissioning (THTR 300).

101 These are normal numbers, as they do occur also with our coal fired power plants - with other words: they are very good for a prototype plant e samTh . e judgemen e drawb n n ca t froe facth m t thamajore th t - ity of unplanned shutdowns happened during experimental transients e numbeth f whico r h exceede e numbeth d f ensuino r g shutdown y farb s .

The experimental program included load following, shutoff of one of o thturbintw e e generator e mototh r fo drives s circulatorsga n , shut- o turbintw e th e f driveo e on n feed-watef o f of r pump switcd an s o t h the stand-by motor driven pump e so-calleth , d fast cool down program e so-calle (noth o fass ts a t d aftercooling mode which functioned excellently froe beginning)th m , load shedding with ensuing station power production, switching between different station power supplies (station transformer, grid, diesel engines), shutoff of one steam generator and putting it back into operation (for this, the load of the other five steam generator . Sevee reduced)b on o o -t s s d ha san , ral experiment e repeateb o t t differend a d ha s t power levels A mino. r number of repetitions resulted from unsuccessful trials.

This experimental program accounte e commissionin r th mos fo f do t g time n additiona I rathe. d ha r e w ,lon g shutdown period, from Nov. 22, 1985 to Jan. 19, 1986, that was necessary for several plant improvement e ensuinth d an gs paper work e activitieTh . s included: Inspection of turbine bearings and control rods, improvement of the fuelling system with respect to hardware (components) and software (computer program), adding thermal insulation to the feed-water and steam lines abov e reactoth e r pressure vessel, cleanine th f o g cooler in the decay heat removing system, exchange of flexible into fixed hydrauli ce feed-wate tubinth r fo gd stea an r m tripping valves, finishing instrumentatio d controan n l etc. Late n shutdowro n times were also used for inspection and minor improvements and sometimes determined the duration of a standstill. The period from Febr. 7 to Febr. 17, 1986 was mainly used for modifications of instrumentation and control, resulting from operating experience.

The second major interruption, costing us another two months, was a consequence of the Chernobyl accident. Thereafter, judging the general climate, we decided by ourselves to stop the experimental program and to go on with the production of electricity at a safe level

102 (40 %). We did so from May 5 to May 27, 1986 with only one interruption (malfunction durin a transienga stea o t m e generatodu t r check- e resumew 7 commissionin2 r ou dy up)Ma n .O g program. Howevery Ma n ,o - suc 0 h3 things always happen lat n Fridayso e , because thee thear y n broadly covere weekenn i d d medi d politicianan a s canno k theias t r staff the minister responsibl e supervisioth r fo e f THTo s caught n wa Rno n to beeing informed about a minor release of aerosols that had happened 2 already on May 4. The release amounted to 1,2 m Ci and added 0,1 Bq/m 2 to the 50 000 Bq/m from Chernobyl that had rained down near THTR one day earlier. These 1,2 m Ci were below the allowed release of m Ci/m Ci/a 2 0 d(1 ) whic alreads i h a veryvaluw lo y e amountino t g about 1/30 of the specific values, per MW installed German LWRs may emit e ministrye stafth Th . f fo e supervisor,th y experts (TÜVd an ) e THTth R safety beel councial n d informeha l d abou e releastth d an e had correctly values unimportanta t i d . Afte e firsth r t excitement, we tried to reach an aggreement with the minister to go on with the operatio n parallei n l with investigatin e incidenth g d improvinan t g plant and procedures. However, political forces were too strongly against such a compromise. So we had a 39 day shutdown. One third of the time was needed for the formulation of requirements which had either been fulfille advancn i d werr o e e fulfilled shortly thereafter. e tims spenth e resewa r gettin Th f fo to g exper- fi t e opinioth d an n nal consen o restart plante th t e trie.W , mako it e t besd th ef to carrying through some changes that also needed the consent of experts and authority.

After restarresumt no d Juln , eto 198di 10 ypowee 6w r productiot a n 80 % as already achieved in April; instead, for about six weeks, we operated the plant at 40 % and raised power to 60 % several times to repeat experimental transients from that level. This became necessary because experienc modificationo t % operatio d 0 e8 le wit e d th hha n s f procedureso e transient additionn e repeateth i b ; f o o t d e dha on s, two more times until it was successful.

The power level of basically 40 % during that period had several rea- - sons. One originated from a near zero availability of the refuelling equipment o defectivt e du , e switch bearings, wrong labellin f pebbo g - les etc. This resulted in a dense core inducing us to risk as little shutdowns as possible and to run the plant at a well-known power level.

103 Another consequence was that we could not change the composition of the core whic s thewa h n characterise y higb d h excess reactivity (enhanced by previous protactinium decay), deeply inserted core rods, and power production shifted to the lower, hot end of the core; in this situation the safest way was to burn down excess reactivity at a reduced power level.

Another reaso r powefo n r reduction, which alone would have been sufficien t leasa t t day-timea t s beet summeha ,ho n r weather that impai- d cooline steare th n mi g pipe rooms abov e reactoth e r pressure vessel. There, according to our own application, temperature margins have been licensed thae lowear t r than necessary fro a safetm y poin f viewo t . Trying to adhere to those margins with additional insulation did not prove successful enough unde l ambienal r t conditions. r Coolinai e th g with cold drinking water ,a simpl e solution s pursue,wa r somfo d e time but was ruled out because of time consuming law of water procedures. Now we have applied for a temporary license to operate with higher temperatures thae stil w enougar t lo lr instruments fo h , leak detection and concrete. The application for the final solution, an air recooling equipment, shall be submitted thereafter. At the end of August weathee th , s turneha r d unusually colde havw o es , ample time to settle that question. Altogether load reductions for the reasons named above cos s severatu l full power days.

Some more days were lost for a formally similar reason, i. e., we had to stick to margins and procedures that we had applied for and that had bee o licenseds n t tha e over-conservative,bu b t o turnet t ou d . This e caseth ,s n wheseverawa o - n l occasions with high excess reactivity and low xenon - we had to build up new xenon after a restart at about % therma 0 3 hala l d powef an r abouday e fo r son t without producing electricity. We shall apply for a change of the license, but the autho- rits alreadha y y indicated that this will onl e granteb y d after extended operating experience.

Please note that lengt f explanatioo h t equivalenno s i n o importancet . o brinT g things back into perspectiv havI e e summarised THTR power production history in table 1.

104 Table 1 Synopsis of THTR power production

days Time reviewed 16.11.198 06.09.1985- 6 294

plant improvement, revision 22.11.1985 - 19.01.1986 58

shutdown ordered by authority 01.06. - 10. 07. 1986 39

shutdown or nuclear power only for other reasons: 69

- instrumentatio d controlan n , including improvements

- leaks and other reasons in the balance of plant

- xenon buildup, inspection, trial shutdown otherd an s s

power production 128 reduced power production after Chernobyl 20 without experiment 27.05.198- . s05 6 minus 2 days shutdown reduced,power,production for various reasons 30 (see text) with experiments 10.07. - 21.08.1986 minus 12 days shutdown or nuclear power only power productio s plannea n d with experiments 78

Summing up, I wish to repeat the valuation given above, that in view of the circumstance e progressth vers i syw achievesatisfactoryno o t p u d .

105 The main reason of success is the faultless function of all essential components, which must not be forgotten over minor problems, however annoying they may be.

Experience with the first large pebble bed reactor

In addition to last year's history, the behaviour of the first large pebble bed reactor will be of special interest:

Core physics are not easy, because there is no incore instrumentation. Generally such an instrumentation is not necessary due to the large difference between operating and limiting temperatures. Only in special situations of the first core as in July 1986 we might wish to e havW hav. e it efairl y good knowledge about reactivity changes with insertion depth f individuao s d groupl- ro rodra d aboud an e san s th t dial distributio f axiallo n y integrated power production from temperature measurements under the core. These values, together with neutron flux measurements and results from pebble flow experiments are put into computer programs and compared to reality, finally ending up with a good understanding of the core behaviour. Computer mode- e corth line f wito g h deeply inserted rods with ensuing shiftf o s flux and pebble flow, which we only will have with the first core, still needs improvement. Even under these conditions short term core calculations can be done with sufficient accuracy.

Fission product retentio vers i n y good e measureTh . d value f noblo s e s releasga e practicallar e y identical with those anticipate y calb d - o influencN culatio e . seeb 3 n g s showca a fi efron n i mn broken pebbles whice a facresul s th i th f o ttha t fission product e mainlar s y retained by the coated particles. 3 There have been more broken pebbles than anticipated n ordea y rb 0 1 , of magnitude out of a total of 675 000. This has to be attributed to harsh experimental conditions: Lack of lubricating gas (NH-J and a very dens eo mant core y du e cor d movementro e s followin experie th g - mental program and due to relatively small amounts of pebbles with- e bottomdrawth t a n. Those conditions wilt prevaino l norman i l l operation e thu.W s expect thae numbeth t f brokero n pebbles wil o bacg l k o valuet s near, zero.

106 i 10H

10-7

8! CO O

10" 11 0,1 i 10 100 HalfUv»[h]

FIG . Calculate3 . measured dan d nobl releass ega e at 40% load (gas exit temperature: 620°C) (THTR 300).

The shutdown system has been amply discussed on German TV. It was said thareflectoe th t r rods cannot shut dowreactoe th n d thar an (the t) do y the core rods alone cannot guarantee long-time cold subcriticality (they eve, do r threno witf theo eo tw hm totally outsid core)e th e . These facts have been confirmed once more by recent jurisdiction. On one occasion, 7 out of 42 core rods failed to reach their full insertion w inchesfe secona depta n y .O b h d move, howeverl al ,o g thed di y the way down. This happened on a unique experimental occasion (den- e cores circulatorl ,al s running adding more density NH,o n d ),an and even then the large reactivity margin already mentioned was not noticeably affected. In normal operation we have not had nor do we expect any trouble.

The fuelling equipment was the source of the small release of May 4, 1986. When switching from manual to automatic operation the gas con- fuee th tenl f entrto y locs release atmosphere wa k th o t d e unfilter- meantimee th n I . , ed operational modes have been change filterd dan s have been backfitted.

A major problem with the fuelling equipment is the fact that the numbe f pebblero s tha n leav core tca th ee decreases abov poweea r e highe Th powere . % th r highe e 0 leve, 4 th e pressur f th ro l e diffe- s circulatorsga e rencth t a e ; this also affect pressure th s e drop

107 fuee e lowermosth alth t f dischargo d ten e tube, wher cooline th e g gas that has come down the inner wall of the discharge penetration changes its way to move up through the downcoming pebbles. Above % power 0 4 e coolin,th blowins ga g g agains e slidinth t g pebbles make increasinglt i s t a yx difficulbo e th r f theo fo o t com t t m ou e e dischargth f o d e en tubee e probleth .Th m wil e solveb l y cuttinb d g some extra holes into that box, allowing part of the gas to search another way e righTh . t siz f theso e e hole s investigatei s mocka n i d- up, the remote cutting is trained in another mock-up. The cutting will be done next year. Until then, it is possible to run the plant at 100 % power on working days and do all the required refuelling at 40 % power on weekends. For this operation mode we need a high availability of the refuelling equipment. This was already achieved on the weekend of Sept. 6/7, 1986, whe operatee nw plane th d80/40/8t a t 0powe% s ra planned. y tha proves sa t ha n feasibilite THT0 ca th dSummin 30 Re w , up gf larg o y e pebbl d reactorbe e s with respecessential al o t l components e natur.Th e e problemth f o st generic no lef s i t ,questioa thee ar y f layouto n o .S n takca e e w full advantag e propertieth e pebblf o eth f - higeso h stability, high fission product retention w temperatur,lo e differen- e betweec s (openinnga fue d windowe an lth g o vert y high temperature applications}, easy handling, and underground storage of spent fuel without or with very little conditioning. The possibility to go to very high temperatur enhances i e a stratifie y b d d coree ,b than ca t easily brought about with the pebble bed by means of a once through core. In addition we can take advantage of on-load refuelling - no burnable poiso t equilibriuma n . highee . i , r fuel utilisation, plus e chancth o react e h high availabilit wits e heava y th h y water reactor f receno s t years. Already with THT e expecRw t short shutdowns r inspectiofo e cominth n i ng years yett ,no , however thin ,i s year. Lookin gy tha e technica sa inte future th tn th o ca d licensin e an ,lw g lesson vere th y o ssatisfactort learnd le s t ha wit THTe 0 yth h 30 R con - R 500 HT e hop.W e eth cep thaf to e wiltw positioa l n i soo e o b nt n star e seconth t d (detail) planning phas f thio e s reactor e plannin.Th g contrac s alreadha t y been finally negotiate d signean d d provisionally. What we now need is a unanimous decision of a large user group. This provided, we hope that the anticipated reorganisation of the user group during the planning phase can be brought about and that the HTR 500 will finall e erectedb y .

108 DESCRIPTION OF CURRENT GCR PLANT DESIGNS (Sessio) nC RESTORATION OF A CEGB MAGNOX REACTOR TO FULL POWER

R.J. PAYNTER, B.J. ROBERTS, P.T. SAWBRIDGE Berkeley Nuclear Laboratories, Central Electricity Generating Board, Berkeley, United Kingdom

Abstract

Hinklee Th y Poin Magno' 'A t x Reacto s commissionerwa 196n i d d 5an recently completed its 21 calendar year of operation. The paper contain descriptiosa operatine th f no g experienc f thieo s plant together with the main technical problems and their solutions.

1. INTRODUCTION The CEGs operateBha d Magnox reactors commercially since the early 1960s. Thes naturae ar e l uranium fuelled reactors cooled by pressurised C0„. This paper examines the operating experience for one of these plants over an extended period. The station in question is the Hinkley Point 'A1 plant which was commissioned in 1965 and has recently completed its 21st calendar year of operation with a mean availability of 73%. A numbe f technicaro l problems have arisen over this extended period of operation which have led to a reduction in the potential electrical generation capacity of the station. In the present paper we show how these problems have been overcome. The strateg s bee yha carefullo t n y monito plane d analysth r an t s it e performance, supporting thi capabilitsD R& wit n ha y which enables a fundamental understanding of the processes involved to be obtained. This has led to the development of whole plant models which offer substantial benefits to the operators in optimising the performance of the plant.

2. STATION DESIGN The Hinkley 'A1 station consists of two magnox reactors housed in spherical steel pressure vessels. Each reactor has six dual pressure boilers, each of which has its own gas circulator, which draws the coolant gas down through the boiler (Figures 1 steae Th m . ansupplie2) d maix ssi n turbo-generators, each with a maximum continuous rating of 93.5 MWe, plus three variable speed turbo- generators of 33MWe which provide electrical power circulatorss ga e fot electricath rne e Th . l generatioe th f no statio s designe nwa e SOOMWeb o t d .

Ill .•.-•':-'" H.P. .SUPERHEATER BANK HP I DRUM} L.P. e SUPERHEATER BANK H.P. L.P EVAPORATOR BANK \ DRUM

H.P.H.T. ECONOMISER BANK

L.P. UPPER EVAPORATOR BANK

L.P. LOWER EVAPORATOR BANK

H.P. TO CIRCULATING OTHER BOILERS PUMP HPLT , LP ECONOMISER 'ECONOMISER CIRCULATING PUMP MVo__ •o—

CIRCULATOR MOTOR

GAS OUTLET

FIGUR 1 ESchemati c Diagra Hinklea f mo 1 Boile'A y r

3. STATION OPERATING HISTORY Whestatioe th n s commissionenwa electricat ne a d l s achievewa e generatioMW d 0 consistentl50 f no y wit maximuha m reactor gas outlet temperature of 373°C, a value 5°C below design level. After four year generation it was recognised that bolts securin steee th g l core support structure were oxidisint ga unacceptably high rate. Lifetime considerations necessitated a limitation on the maximum reactor gas outlet temperature of 360°C. At about this time failure of one of the main turbines occurred. It was recognised that this was a generic fault, which required modifications to be made to all six of the main turbines over the following four years. This constrained the station to a reduced power output over this period with the boilers operating under 'off-design* conditions.

112 FIGURE 2 Sectional View of a Hinkley 'A' Boiler

113 When all the turbine modifications had been made the reactors were returned to full power, albeit with the restriction on gas outlet temperatures dictated by oxidation. It gradually became apparent tha beed t ha gradua na l deterioration ni electrical generation capability of 30 MW to 40 MW, when one compared reactor performance pre and post the period of low power operationt firsa s t wa assume t I . d that thi a s par swa f o t station ageing process. However by the later 1970s it was recognised thadecline th t performancn i e s greateewa r than would be expecte n theso d e grounds. Plant monitoring showed than a t increase in boiler gas outlet temperature of between 9 - 10°C was observed r operatio,fo constana t na t circulator speed. This suggested that the overall boiler heat transfer coefficient had been reduced. Detailed analysis of station performance during steady state operation, confirmed that this consistenswa t t ne wite th h station generation capabilit dropped yha approximately b d W M 0 4 y (Figure 3). In order to do this it was necessary to normalise the generation in 1965 and 1975 to a fixed set of boundary conditions via a steady state whole plant model.

: VALUE VALUE PARAMETER 196N I 5 NORMALISED 197N I 5 NORMALISED

Boiler gas inlet temperature (°C) 371 365 360 365

Condenser pressure (Pa) 3000 3500 3300 3500

Circulator speed (rpm) 2840 2700 2650 2700 Boiler gas outlet temperature (°C) 181 175 182 183 Net electrical generation (MWe) 518 490 432 450

FIGUR E3 Normalise d Operating Parameter Comparison 196 d 1975an 5

. 4 EXAMINATIO BOILEF NO R HEAT TRANSFER SURFACES This resulted in detailed examination of the boiler heat transfer surfaces. Whilst on the gas side a thin black uniform deposi s observedwa t , examinatio waterside th f no e revealee th d presence of substantial magnetite deposition. This was concentrate n unheatei d d section boilee th f sro tube, i.e. outside the boiler ducts. However after acid cleaning only a small improvement in station performance was observed. Microscopic examinatio s thu nwa s sid sga e carrien o t ou d heat transfer surfaces studn ,o s removed froe finneth m d boiler tubes. A special tool was built to do this without distribing the deposit. A carbonaceous deposit was observed to cover most e . tubth Mm e f 0 o surface20 d depta an o betweef t sm o h p 0 10 n

114 Deposit thicknes measures swa d usinga scanning electron microscope.

BROKEN WELD DEPOSIT TO TUBE

STRIPPED AREA

Average mean thickness showe b no t about 125;um.

ExaminatioM FIGURSE E4 Deposif no t Structure

Figure 4 shows the deposit on a stud examined by SEM. Compositional analysis showed it to be mostly carbon (by volume), bumassy b t , approximatel consiste2 50 y variouf o d s iron oxides. The deposits also formed in plumes which grew vertically out of the surface (Figure 5). The volume of the plumes was approximately 10% of the deposit envelope and the plumes themselve porosities sha approximatelf so y overale 902Th . l density of material within the deposit envelope was calculated to be approximately 25Kg m . No evidenc s founewa d withi deposie th n substratr to r fo e catalytic growth. What evidence there is, suggests that the deposit forms homogenously withipolymerisatioo t cor e e nth edu n of CO under irradiation. It also appears that deposit formation requires low hydrogen levels in the gas. The Hinkley A gas circulator bearingr ai d d thusha circuie san sth d lowetha r hydrogen levels than other magnox reactors, which had oil lubricated circulator bearings.

5. THEORETICAL CONSIDERATIONS OF DEPOSIT HEAT TRANSFER A theoretical assessmen effecte th f coolanto f so t flow throug deposie hth t suggested: deposie forcee Th th n so t ) wer(1 e unlikel affeco yt e th t plume orientatio morphologr no y

115 Higher magnification using transmission electron microscopy showed tha columne tth s were porous and composed of microspheres.

x7840 X147000

Overall deposit envelope porosity show aroune b no t d 99%.

FIGURE 5 Transmission Electron Microscopy of Deposit Morphology

s floga w e (iirateTh ) s woul e sufficientlb d w througlo y h the deposit to be essentially laminar. s thereforIwa t e suggested thae primarth t y e effecth f o t deposit would be to extend the fluid boundary layer adjacent to the surface (Figure 6). Paynter and Roberts [1] developed a

BOILE/ // R TUBE/STUD f———I LAMINAR SUB-LAYER 7——;- DEPOSIT

FIGUR 6 ELamina rTube th Floe t Surfaca w r Clea fo d Depositee an n d Tubes

116 simple model on this hypothesis. Gas flow in a boiler tube during normal operatio essentialls ni y turbulent, wit hthia n boundary layer of laminar flow. In a turbulent oagnox reactor during normal operation this layer is approximately 200 pm thick. Paynte d Robertsran ' calculation assumptioe base] th [1 sn o d n that the deposit essentially extended the laminar layer concluded that a 402 reduction in film coefficient of heat transfer. This implied that the effective thermal conductivity of the deposit was 0.05 W/m/°K. If thes w thermalo e l conductivities coul e substantiatedb d , It was clear that most of the station loss of generation capacity resulted from these gas side boiler deposits.

. 6 EXPERIMENTAL VERIFICATIO THERMAW LO E LTH CONDUCTIVITF NO Y OF THE DEPOSIT Direct thermal conductivity measurements were ruled out becaus doubtf o e s concernin e relevancth g measurementf o e s made n deposio n stagnani t t rather than flowing gas. Consequentlya comparative measurement on clean and deposited tube was made under representative boiler operating conditions (Figur. e7) useg n electricallri a d e Th y heated tube 0.3 lengthn mi , which had been removed from the reactor. A pressurised rig meant that the correct rang Reynoldf eo s numbers coul obtainee b d d without excessiv velocitiess ega , which could have damage r removeo d e th d deposit.

A pressurised rig was constructed to measure ip^&^jpjL ^J fêSTA 1^>^ ELECTRICAL the deposit i \H^^^J^r/ yC*XS^JS2, HEATER

heat transfer t resistance.

TUBE UNDER TEST

FIGUR 7 EExperimenta l Measuremen Deposif o t t Thermal Conductivity

117 The results are shown in terms of the dimensionless heat transfer co-efficient (Nussel Reynolds ga e tth snumberd an f )Nu numbe Prandtd an r l number) Pr R ( s

66 ^- 0.186 xR°' m Pr1/3 In Figure 7 regression lines are shown for the clean and deposited tubes. The regression line for the deposited tube suggested a boundary layer extended by 85 pm (in terms of Paynter and Roberts' theory) compared to an actual measured deposit thickness of 110 urn. This was felt to be a a reasonable agreement, since the surface of the deposit was rougher than the clean tube and some conduction must occur through the deposit plumes t leasA . t gavi t e sufficient confidenc deploo t e e th y necessary resources on deposit removal techniques.

7. DEPOSIT REMOVAL It was obviously economic to devise a cleaning method which requirt dino d e remova boilere th f lo s from service. Laboratory experiments on air, CO and oxygen dosed coolants suggested that oxygen dosing was the most effective way of removing the deposit by oxidation. Sufficient oxygen has to be introduced to oxidise any free CO in the coolant before deposit oxidation occurred. The deposit oxidation was very temperature sensitive (Figure 8) thus removal from the upper boiler superheater banks was expected dayw tofe s occu a wherea n i r économisee sth r sections, operating at 130°C, were expecte tako t d e many months (Figur. e9)

TIME TO -BURN OFF 98% OF CARBONACEOUS DEPOSIT SOOvpmOt/COj

TIME.OAYS FIGURE 8 Effect of Temperature on Deposit Burn-off in 900 vpm 0_ and C0_

The oxidation essentially removed the graphitic part of the deposit and freed the iron oxide particles, which were gradually remove y filterb coolande th n i st treatment by-pass circuit. Howeve smalra l increas activite th n ei y aroun circuie th d s twa measured. This indicated that there had been significant deposition in the core, which when removed by oxidation of the carbon released activated iron oxide particles.

118 TIME DEPENDENCE Of DEPOSIT REMOVAL 190 H P EVAPORATOfl BANK BOILER 6. REACTOR 3

HO-

200 250 0 10 TIME (DAYS° 15 )

FIGUR E9 Predicte d Burn-Of Deposif o f t from Different Section Boilee th f rso

A comprehensive safety assessment of the effects of the oxygen injectio n fueno l performance, graphite moderator lifetime and the performance other key reactor components, showed that the proposed metho s feasibldwa d safeean . Oxygen injection commenced in reactor between January and May 1985. Plant monitoring showed that within days deposit removal commended in the higher temperature regionboilere th f so . Thi s evidenceswa d by a rapid change in superheater steam outlet temperatures, followe progressiva y b d e increas P steaH n mi e water flow (Figure 10). However the LP water/steam flow initially decreased and overall station electrical generatio s littlnwa e improved. s thoughIi t t that deposit removed from upper banks initially deposite lowee th r n s o inledbank ga d alste san th otemperatur e to the lower LP banks reduced, thus limiting the LP water/steam flow rate. Only when substantial cleaning of the lower banks was effected did total water/steam flow rates, and hence electrical generation increase. Table 1 shows how boiler parameters have improved following the oxygen injection. Overal e electricath l l generation capacity statioe oth f bees ha nn increase y approximatelb d MWe0 2 y , witha benefit to the CEGB of approximately £2 x 10 per annum in replacement fossil fuel costs. Recent s outlerelaxatioga e tth temperaturn i n e restriction duoxidatioo t e n limit 370°Co t s , together wite improveth h d boiler performance, has contributed to making Hinkley Point 'A' the United Kingdom's leading Magnox station d fift,an h position in the world in terms of culminative electrical generation. Thus in 1984/85 the station generated more electricity than in all but previous it f o so yeartw operationf so . Therpossibilita s i e y tha n 1986/8 i ty possibl ma t 7i y challeng annuas it e l generation e 21srecorth t n yeai d operationf o r .

119 5 HP 4 oc «" 3 > US J

LUC/) LP Oj£ ZLU

B FE N JA MAR APR MAY

y 10

ZLU OIUuJ R JAAP N R FEMA B MAY

TIME DEPENDENC CHANGF EO STATIOEN I N PERFORMANCE DURING OXYGEN INJECTION INTO GAS CIRCUITS OF REACTOR 1

FIGURE 10 Progressive Changes in Boiler HP and LP Water/Steam Flow and Station Electrical Generation During Oxygen Injection

TABL 1 E CHANG N BOILEI E STATIO D AN R N PERFORMANCE DUE TO OXYGEN INJECTION INTO THE REACTO S CIRCUITGA R S

VALUE PRIOR VALUE POST PARAMETER TO OXYGEN OXYGEN INJECTION INJECTION Boiler Gas Inlet Temperature CO 370.0 370.0 Boile s OutleGa r t Temperature CO 184. 1 178.7 Boiler Gas Flow Ikq/s) 787 800 Feed Water Temperature CO 68.5 70.0 HP Steam Temperature CO 359.8 361.9 LP Steam Temperature ( "CI 344.8 348.3 Boiler HP Water/Steam Flow (kg/ s) 32.7 36.4 Boiler LP Water/Steam Flow (Kg/s) 21.8 21.0 Ratio of HP and LP Flows (-J 1.50 1.73 , LP Drum Pressure (N/m1! 13.3x10 13.5x.10 HP Drum Pressure (N/m'l 43.1x10 45.4x10 Station Net Electrical Generation at a Condenser (MWc) 458 479 Pressure of 3386 N/mJ

The values quote e average e ar statiod th r fo sn

120 ACKNOWLEDGEMENTS This pape publishes i r permissioe th e y b dth f no Central Electricity Generating Board. The authors wish to express thanks to all their colleagues at Hinkley Point Powe Soute rth Statiohn i Wes d tnan Region Scientific Services Department who took part in this project.

121 THE MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR (MHTGR)

A.J. NEYLAN GA Technologies, Inc., San Diego, California, United State f Americo s a

Abstract advancen MHTGe a Th s i R d reactor concept being developen i d the USA under a cooperative program involving the US Government e nucleath , r industre utilitiesth d e an y Th . design utilizes basic HTGR feature ceramif so c fuel, helium coolan a graphit d an t e moderator. Howeve e specifith r c sizd configuratioan e s selectei n o utilizt d e th e inherently safe characteristics associated with these standard features coupled with passive safety systems to provid a esignificantl y higher margi f safeto n d an y investment protection than current generation reactors. Evacuatio r shelterino n t requirede publino th s f i co g . The major component e nucleath f o sr steam supply, with special emphasis on the core, are described. Safety assessment concepe th discussede f so tar .

The Modular High-Temperature Gas-Cooled Reactor (MHTGR beins )i g develope e Uniteth n di d States unde a cooperativr e program involving the U.S. Government's Department of Energy (DOE); the nuclear reacto rA Technologie G industr y b d le y s Inc. (GA)e ,th industrial pioneee HTGth R f o rconcepte utility/use th d an ; r industry represented by Gas-Cooled Reactor Associates (GCRA). The design is an advanced version of the basic High Temperature Gas-Cooled Reactor (HTGR) concept. The overall goal is to provide safe economical power with a standardized, pre-licensed factory fabricated reactor.

This modular design uses the following standard HTGR features: o refractory coated particle fuel capabl f retainino e g fission products at very high temperatures,

123 o graphite moderator which remains stable to very high temperature higa s h ha hea d tan s capacityd ,an

o helium coolant whic s inerti h , non-corrosived an , remains as a gas under all operating conditions.

n previoui s A s concepts these standard HTGR features provide th e capability to operate at temperature for high efficiency electricity generation and offer the potential to access the cogeneration and high temperature process heat markets. The basic characteristics also provide inherent advantages associated with safety, investment protection d environmentaan , l compatibility which are greatly enhanced by the special design features of the MHTGR reactors.

DESIGN DESCRIPTION:

w thermaMHTGRe lo Ith n e l,th output annulae ,th r geometre th f o y a totall f o e y us corepassiv e th , e ultimate heat removal system e installatioanth d e reactoth f o nr syste beloa n i mw grade silo are design features incorporate desige th provido n nt i d e safety which is not dependent on active safety systems, on operator e evacuatioth action n o r r shelterino so n publice th f o g.

The reference MHTGR plant design consist f fouo s r reactor modules, each rated at 350 MW(t), coupled to two steam turbine generators yielding a net power output of about 550 MW(e). An added advantage of the modular concept is the flexibility of providing sequential deployment of individual reactor modules to match the end user's load growth requirements or financing constraints.

124 e four-modulTh e plan s dividei t d into majotw o r arease th , Nuclear Island comprising the reactor enclosures, reactor modulese safetth d y an ,relate d systemsbalance e th th d f o e,an plant housing the electric power generating systems. Each reactor modul s housei ea vertica n i d l cylindrical concrete enclosure whic s fulli h y embedde e earthth n .i d Each reactor enclosure serven independena s a s t confinement structure havina g vente d filterean d d exhaust system e NucleaTh . r Island portion of the plant also includes auxiliary structures which house common systems for fuel handling, helium processing, and other essential services.

The modular reactor components are contained within three steel vessels: a reactor vessel, a steam generator/circulator vessel, and a connecting concentric crossduct vessel. The vessels use existing LWR technology and have weight and dimensions within previously demonstrated fabrication and transportation limits.

The reactor vessel, 6.9m (22.5 ft.) in diameter and 22m (72 ft.) in length, contains the core, reflector, and associated supports. A shutdown heat exchange a shutdow d an r n cooling circulatoe ar r located at the bottom of the reactor vessel. Eighteen top- mounted standpipes contai e controth n d drivro l e mechanismd an s hoppers containing boron carbide pellets for reserve shutdown. f theso x eSi standpipe e alse accesar sth o s port r refuelinfo s g and in-reactor inspection.

A helical coil steam generato electrie th d an rc motor-driven main helium circulator are contained in the steam generator vessel whic s 4.3i h 4 5 ft.ft.m(1 (8 n diamete)n lengthi m )i 26 d .an r Feedwater enters the steam generator through a bottom-mounted

125 heade d superheateran d steam exits throug side-mountea h d nozzle. The steam generator is similar to/ but simpler than, the Fort St. Vrain (FSV) design. The main helium circulator is located above the steam generator.

The reference design circulator utilizes magnetic bearings and a submerged motor to eliminate the water ingress problems that have plagued Fort St. Vrain (FSV). Although this is an important and preferred design selectio o meet n t availability goals woult i , d als e possiblb o meeo desige t e th t n requirements wit greatla h y simplified third generation water bearing system that has already been full scale tested at GA. The single stage axial circulator, driven by a variable speed 3.6 MW electric motor, delivers 158.0 Kg/s (1,254,000 Ib/hr) of helium at 6.378 MPa (925 psia.) primare Ith n y flow circuit heliue ,th m coolant t approximatel,a y 260°C (500°F), flows downward throug e coreth h , s wheri t i e e nucleaheateth y b dr reactions t heliumho e t approxia ,Th . - mately 688°C (1270°F), then flows through the inner cross-duct and downward over the steam generator bundle where its heat is transferred to the water to make steam. The cooled gas then flows upwarannulun a n i d s betwee e steanth m generator bundld an e the vessel recompresses ,i circulatoe th y db drived ran n inte oth annulus betwee e inne th d noute an r r crossduct e coos Th ga l. entering the reactor vessel flows up through channels in the lateral core restraint structure to the top of the core to complete the circuit. All surfaces of the steel pressure vessels arcontacn ei t with heliulowee th rt m a temperature .

In the secondary flow circuit, feedwater is transformed to superheate de once-through th stea n i m , uphill boiling steam

126 generator e steaTh .m generator delivers 137.3 Kg/s (1,089,700 lbs/hr.) of superheated steam at 17.340 MPa (2515 psia) and 540°C (1005°F two-turbina o )t e generator unit. Exhaust steam froe mth turbine is condensed and returned to the steam generator through a standard condensate/feedwater system. Control of operations for the complete plant is performed from a single control room in the control building located adjacent to one end of the reactor services building.

A completely independent shutdown cooling loop, comprising a heat exchanger and circulator, is provided at the bottom of the reactor vessel for maintenance purposes. Provision of this capabilit s directli y n responsi y o uset e r requiremento t s minimize downtim meed ean t availabilit yshutdowe th goals n I n. cooling mode hot helium enters the shutdown helium to water heat exchanger throug a hflow-activate d shut-off valve. After transferrin s heat it e gcoo th ,l heliu s compressei m e th y b d shutdown circulator e compresseTh . d heliu s thei m n discharged into a plenum at the bottom of the reactor vessel and returns to core th e f througo p th annulae to e hth r flow passage betweee nth core and the reactor vessel.

MHTGe th Ro t performanc Thy eke capabilitd ean refractore th s i y y coated fuel. As in the FSV fuel, fissile and fertile fuel particles are blended into 1.27cms (1/2") dia x 5.08cms (2") long fuel rods which are in turn loaded into prismatic fuel elements approximately 78cms (31H) in height by 35.6cms (14") across flats. Vertical coolant holes are provided in the fuel element blocks. Major differences from the FSV fuel are the adoption of Low Enriched Uranium (^19.9%) to meet National policy guidance on proliferation resistanc d improvementean e fueth ln i smanufactur -

127 ing process to produce a fifty fold improvement in fuel quality remarkabla — — e objective considerin excellene th g t operating experienc performancd ean fuef eo FSVn i l .

The fuel elements are stacked in columns to make up an annular- shaped core havin a radiag l thicknes f abouo s t 0.9 m3 ft.)( n a , average outer diameter of 3.5m (11.5 ft.) and a height of 7.6m (25 ft.) Unfueled graphite blocks surroun e activth d e coro t e form replaceable inner and outer radial and upper and lower axial reflectors. Permanent reflector blocks are located at the outer peripher replaceable th f o y e graphite blocks.

The total assembl s restrainei y a stee n i dl core barrel within the reactor pressure vessel. Lessons learned at FSV on preventing core fluctuations have been factored inte designth o . Althoug e cor th hs talleei r both end f eaco s h colum e pinnear n d (th ecore t FSVdesigth a e d x )pressuran fi n e drop (the driving force s considerabl)i e eliminatioth y o t lowe e f orificdu ro n e valves and the six year core region fuel loading used on FSV.

Refueling of a module is accomplished with the reactor shutdown and depressurized. After removae requireth f o l d control rods and drive mechanism n auxiliara y b s y service machinee th , refueling machine is located over the desired refueling port by the crane in the reactor services building and fuel elements are removed from the core and transported in shielded casks to the fuel storage area in the fuel building. One half of the fuel is replaced each 20 months resulting in a 3.3 year fuel cycle.

o independenTw d diversan t e reactivity control systeme ar s provided. Reactor powe s controllei r oute2 1 inne6 d ry b dan r

128 articulated control rods that trave vertican li l channels located e inne ith d outen an r r graphite reflector regions. Twelve reserve shutdown channel providee sar d which, when activated, permits small boron carbide spheres to enter channels located in the innermost row of fuel columns.

The annular core geometry and power density of 5.91 w/cc were selected to limit peak temperatures and hence fuel failure during loss of forced circulation events such that the user requirement t exceedinono f S EnvironmentagU l Protection Agency's Protective Action Guidelines (PAG) dose limits at the site boundary would be achieved.

Decay heat removal during shutdown, inspectio r maintenanco n e periods can be accomplished using the main heat transport loop wite steath h m generate e secondarth n i d y circuit bypassine th g turbin condensind an e condensere th n i gdecae Th .y hea n alstca o be removed by the shutdown cooling system located at the bottom reactoe oth f r vessel.

Io nindependen tw additio e th o nt t active systems use o removt d e heat from the core i.e. the main loop (steam generator/ circulator) and the shutdown cooling loop (heat exchanger/ circulator a thir) d totally passive syste s locatee i m th n i d reactor cavity adjacen e reactoth o t tr vessel. This system, designate e Reactoth d r Cavity Cooling System (RCCS), removes radiant heat from the reactor vessel under normal and accident conditions using naturally circulating ambient air in specially designed cooling panels. The system is totally passive; no valves, fan mechanismr o s e onl e utilizedth yar ss i syste t I .m neede o removt d e heat froe cor meeth o mt e t regulatory safety requirements.

129 SAFETY ASSESSMENT;

The safety characteristics of the MHTGR are a result of the abilit o effectivelt y y utiliz inere th e t coolant capabilite ;th y refractore oth f y coated fue withstano lt d high temperaturesd ;an the high thermal inertia and high temperature stability inherent in the graphite core and support structures. These unique features have been exploited in the MHTGR to yield a reactor that does not depend on active engineered safety features or human actions for safety of the public or the investment.

Helium gas e reactoth , r coolant s ineri ,o n o thers te ar e chemical reactions betwee e fuele coolanth th n ,d an graphitt e core structure r reactoo , r vessel which could lea o firet d r o s explosions. Further, e nucleao effecheliun th s n o ha tmr reaction.

The graphite fuel element d reactoan s r internals which makp u e the reactor core have a high heat capacity and maintain strength to temperatures beyond 2760°C (5000°F) resulta s A . , temperature changee corth e f o soccu r slowl withoud an y t core damagth e o t e structure even in the event of accidents where the capability to remove heat is impaired.

As stated previousl e annulath y r core geometry, core power density, and the total module power of the MHTGR have been chosen such that the decay heat generated within the core can be passively removed by means of conduction, radiation, and natural convection without the fuel reaching a temperature at which the refractory coating would fail during an accident.

130 Primary cooling systems and shutdown cooling systems are available to remove heat and the core temperatures do not exceed normal levels. Even if both active cooling systems are unavailable, decay heat is dissipated by conduction and radiation e reactotth o r cavity cooling system e reacto(RCCSth n i )r enclosure. The heat transfer is sufficient to avoid damage to the fuel or reactor components. The maximum fuel temperature is limite o aboudt t 1600°C, well belo failure wth e temperature.

The high temperature stability and slow heatup characteristics of the graphite reactor core and fuel not only retains fission products within the coated particles but provide a user friendly and forgiving reactor design that allows many many hourr fo s operator corrective action.

The inherent characteristic MHTGe th Rf o sals o assur e safeteth y e systeoth f m against other type accidentsf so .

o Anticipated Transients Without Scram (ATWSe ar ) mitigate e stronth y gb d negative temperature coefficien d zertan o coolant void coefficiene th f to reactor core. These inherent characteristics cause the reactor to automatically shutdown in the event reactor temperature rises above normal as a result of equipment failur operator eo r error. e reactivitTh o y control systems provide considerable margin over the most adverse reactivity insertion accident including unlimited water ingress.

131 o Chemical reactions between steam and graphite can only occur as a result of accidents or equipment failures allowin e steath gr wateo m o comt r e into contact with the graphite. At normal operating temperature reactioe sth n rat s insignificanti e e Th . reaction rate increases with temperature but, in any event s highli , y endothermic (heat absorbingd an ) therefore self terminating.

o Air/oxygen reactions with graphite are exothermic (heat releasing) but can only be sustained if multiple breache e reactoth f so r vessel f i occu d an r e graphitth s heatei e o temperaturet d s much higher than normal. The silo installation limits the air available to support combustion in the extent of any reaction woul minore b d . Even were graphiteth o t e e burneb d awae fissioth y n product activits i y retaine refractore th y db y particle coatings.

No public evacuation or sheltering is required even for severe low probability accidents becaus consequencee th e mitigatee ar s d e inherenbth y t passive feature MHTGRe th f .o s

CONCLUSION:

The Modular High Temperature Gas-Cooled Reactor (MHTGR) is a second generation nuclear power system which can satisfy the concern e publicth f e governmento sth , e utilitiee th , th d an s investor community about nuclear safet d investmenan y t protection. Based on technology developed and demonstrated in the U.S. and Germany, the unique system makes use of refractory

132 coated nuclear fuel n inera , s heliuta coolans ga m d graphitan t e aa stabls e core structural material e safet Th protectiod . an y n of the plant investment is provided by inherent and passive features and is not dependent upon operator actions or the activatio f engineereo n d systems e higTh h. performance MHTGR provides flexibilit n powei y r outpu d sitingan t , competitive energy costs, and can serve diverse energy needs both domestically and internationally.

The work described in this paper reflects the combined efforts of all the US program participants under contract to the Department of Energy I wis o .exprest h y appreciatiom r s fo E DO o t n permission to publish this paper.

ACKNOWLEDGEMENTS

The author woul . MillunzidC lik. A o thant e. ,kMr Acting Director, Divisio HTGRf o n , U.S. Departmen Energyf o t r ,fo approva o publist l h this paper. Thank alse sar o expressed managemene tth o A TechnologieG f o t s r permissioIncfo . o nt write and present this work, which was supported by the Department of Energy, San Francisco Operations Office Contract DE-AC03-84SF11963.

133 STATU GAS-COOLEF SO D REACTOR DEVELOPMEN USSE TH R N TI

V.N. GREBENNIK I.V. Kurchatov Institute of Atomic Energy, Moscow, Union of Soviet Socialist Republics

Abstract

The paper contains a description of HTGR concepts as being developed USSe r electricitith nfo R y generatio d procesan n s stea d heaan m t production. The HTGR-M, a module type reactor, and the VG-400 are presented together with their main design features and their applications.

In the USSR the nuclear energy is currently developed predo- minantly in the direction of electricity generation where it com- petes successfully with fossil-fuell power plants. Efforts are also being made for utilization of nuclear energy sources for low-temperature heat production on the basis of nuclear district heating plants (IfDHP) and nuclear co-generation plants (NOGfP). For these purposes use of commercial light water reactors. At the same time more actual is becoming the problem of ex- tending field nucleaf so r energy applicatio industriar nfo - di d lan strict heating and industrial processes consuming great amounts of oil and gas. In the conditions of large-scale development of the nuclear energy ensurin highef go r nuclea radiatiod ran n safetd yan further improvement of operation reliability of the nuclear energy sources is a task of great importance. To a significant degree these requirements may be met on account of realization of a new nuclear energy direction introducti- high-temperature th of no e gas-cooled reactors (HTGR) whose advantageous features are high nuclear and radiation safety, effective utilization of the nuclear fuel and capability to generate high- temperature heat.

135 These features are accounted for by use of the helium coolant which is chemically inert and has no phase transitions, and use of graphite as the structural material for the core. In industry HTGR can be used for co-generation of steam, high- of methane temperature head electricitytan steae th ms . A conversion/an d other high-temperature processes are being mastered on the basis of power from HTGR ,possibilita y will ope wid r d effectivnfo ean e application of the nuclear energy for substitution of oil and gas in various power-consuming industry fields. analysie Th scale th d structur ean f so powef o e r consumption industre th n i y shows that there both powerful plants (about 1000 MW(th) and higher) for power supply of large industrial centers and units of relatively low power capacity can find application /1/. In the USSR R and D works are being carried out on some pilot scale plants of various capacities as well as on application of HTGRs for process steam and heat supply. The HTGR-M reacto modulaf o r r type wit hunia t powe 200-30f ro 0 MW(th) is considered as a small pilot plant. Taking into account the experience gained earlier, development of the HTGR-M reactor is based on the following main statements» - use of a pebble-bed core with spherical fuel elements and realizatio variouf no s principle theif so r movemen reace th -n i t tor; - compensation for burnup excess reactivity by spherical ab- sorbers; controe th f d safeto lan e yus rod- side s th onle n reflecyi - tor; ensurin- higf go h safet accounn negative yo th f o t e temperature coefficient of reactivity, no phase transitions and chemically inert coolant;

136 currene th f o t e technologus - manufacturinf o y g metal pressure vessel applicatiod san materialf no s availabl reacton ei r building. The HTBR-M reactor uni designes i t verificatior fo d basif no c technical solution technologd san workf yo HTGn so wels Ra s la experimentar fo l test high-temperaturf so e heat, stea d electriman - city production processes. The HTGR-M core is installed inside a reinforced steel vessel about 5 m in diameter. The primary circui reactoe designee b th vero n f tw to rca -n i d sions: with the steam generator or with the steam reformer. The HTGR- M with the steam generator can be realized at the first stage since it proposes a lower helium temperature (750°C) and a simpler heat exchange equipment e secon.Th d version proposes testin steaf go m conversio methanef no , wit reactoe hth r unit operate temperatura t a d e as high as 950°C. importann A t facto realizatior rfo modular-type th f no e plant is unificatio d commercializationan standarf no d equipment. Large industrial plants (about 1000 MW(th) can be created by integration of several one-type modular low-power units n thi.I s casprobe th e - lem redundancf so d reliablyan e power suppl continuouf o y s technological productions woul favorable db y solved. The pilot industrial-plant VG-400 is designed for co-generation of high-temperatur 950°C)eo t heap (u t, stea d electricity,whicman h wil usele b varioun enabldi o t t ei s power-consuming industries such as chemical, petrochemical, metallurgical, etc.). course Inth VG-40f eo 0 realizatio desige nth d technicanan l solution maie th n f sequipmeno d system tan industrian a f so l plant with the HTGR reactor will be confirmed and tested in the scale of industrial experiments. In accordance with the objective of the plant the ITPH schemes and parameters were chosen, which enables attainment of characteristics suitable for transition to an extensive and effective indust-

137 rial utilizatio nucleae th f no r energ procesr yfo s stea head man t production in industry. Sinc plane pilot-scala eth s i t natures desigs i it y H eb -,UP ne thao ratee s d th t d parameter reachee sar d step-by-step with continuous rise in the reactor temperature and power levels and simultaneous tests of the steam generating and heat power equipment, core and high-temperature energotechnological station. Two schemes of the reactor unit were considered: conventional one wit heae hth t exchanger, steam generato blowes ga d rr an connec - ted in series, and proposed one with bypass valves in the heat exchange loops (bypass scheme) /2/. The followinreactoe th s rha g main parameters: Reactor power 1060 MW(th) Helium pressure in the primary circuit 4.9 MPa Helium temperature at the core outlet 950°C Helium temperature at the core inlet 350°C For the VG-400 reactor the integrated arrangement of the primary circuit equipment installed insid prestressee eth d concrete many-chamber reactor vessel has been adopted /3/« The reactor unit has the intermediate helium circuit which prevents penetratio radioactivf no e fission products intproe oth - cess circuit and contamination of the primary circuit by products of the technological process. The intermediate circuit increases the unit safet maked yan flexiblt si applicatioo et variouf no s techrio- logies. One of the most important and strained elements of the intermediate circuit is the high-temperature intermediate heat exchanger (IHE) which serves for transfer of high-temperature heat from the reactor to the process circuit. The VG-400 reactor is equipped with cassette-typa e heat exchanger whic installes hi d insid cyline eth - drical prestrestee cavitth f yo d concrete reactor vessel heae .Th t exchange designes ri normar dfo l operatio 950-750°Ct na . core once-through-then-oue th eth n I t (OTTO) principl sphef o e - rical fuel element movemen realizeds i t , which will permi fuee th tl

138 element temperature to be reduced and the refuelling to be performed with the reactor in operation. The core design with a system of discharge holes and radially homogeneous loading of fuel elements bees ha n adoptebasie th cs a dversion . Practically all components of the main equipment of the plant are prototype, which enables, basing on the technical solution on pilot plant VG-400 advance HTGR-equippee ,th th o et d plant variouf so s capacities and objectives. A full-scale development of the nuclear energy will require to solve the problem of reproduction of the nuclear fuel resources. The analysis shows that this problem coul solvee b d e meany th b d f so fast breeders which have a high rate of excess plutonium buildup and short doubling time. In the USSR investigations on the helium breeder, which is considered an alternative of the sodium-cooled breeder, are in progress. Application of HTGR as an effective higher-safety nuclear so- generatiow ne e urcth f eno importann wila e lb t facto powen ri r supply of the national economy and in saving scarce fossil fuels.

REFERENCES

1. V.N.Grebennik "Status of Gas-Cooled Reactor Development in the USSRV Paper presente I InternationaVI o dt l Conferenc Gasn eo - Cooled Reactors", PRG, Dortmund, 17-18 September 1985» 2. F.M.Mitenkov, et al. "Nuclear Bnergo-Technological Plant VG-400". Voprosy atonmoi nauk tekhnikiii . Ser. Atomno-vodorodnaya energetik tekhnologiya"ai , 1985p.59-66.(I, 2 ,N n Russian). 3. U.N.Ponomarev-Stepnoi, et al. "Designs of prestressed concrete reactor nationavesselw ne r sfo l HTGRs" ibid. 1984 2(18),N , p.30-32 (In Russian).

139 THE NUCLEAR ELECTRICITY AND HEAT GENERATION USING THE VG-400 REACTOR

V.V. BULYGIN, V.E. VORONTSOV, V.N. GREBENNIK, E.M. IONOV, A.N. PROTSENKO, A.Ya. STOLYAREVSKU, S.V. SHISHKIN, Yu.K. PANOV I.V. Kurchatov Institut f Atomieo c Energy, Moscow, Unio f Sovieno t Socialist Republics

Abstract

The paper contains a description of HTGR concepts as being developed USS e r electricitith nfo R y generatio procesd an n s stea d heaan mt production HTGR-Me Th . modul,a e type reactorVG-40e th e d 0ar ,an presented together with their main design features and their applications.

At the present stage of the nuclear energy (NE) development more actua becomins li problee gth expandinf mo g field reacf so - tor energy application for industrial and district heating and industriar fo l processes consuming larg natud e an amountl - oi f so ral gas. Against this background particularly important is further improvemen nucleae th f d radiatioo tran n nuce safetth -f o y lear energy sources» These problems could be solved to a significant degree by introductio high-temperature th f no e helium-cooled reactors which are currently under development in the USSR. One of the nuclear designs which represents this new nuclear energy direction is the VG-400 high-temperature gas-cooled reactor. The pilot energotechnological plant wit VG-40e hth 0 reactor is intended for co-generation of high-temperature heat, process steam and electricity for chemical, pertrochemical and other industries /1/. Schematic diagra plane th shows i tf m o Pig.1n ni t consist.I s of the reactor unit (RU), steam turbine unit (STU) and chemical technological production (CTP) where hea supplies ti d througe hth high-temperature intermediate heat exchangers (IHE) /2/.

141 I ü^~ ^ ~X_ VÎJ L

FIG. Schemati1 . c diagra powef o m r technological plant using VG-400 reactor.

1. VG-400 reactor 2. High-temperature intermediate heat exchanger (IHB) Stea. 3 m generator 4* Gas-blowe primare th f ro y circuit Bypas. 5 s valves 6. Conversion apparatus Steam-ga. 7 s mixture preheater 8. Steam generator bloweintermediate s th Ga f . ro 9 e circuit 10.Turbogenerator 11.Separator-steam-preheater 12.Condenaate pump 14.Low-pressure preheater 15.Deaerator 16.Feed pump 17.High-pressure preheater

142 Creation of the VG-400-equipped plant will permit experience of designing, manufacturin d operatingan helium-coolee gth d reactors designe variour dfo s energ technologicad yan l purposee b o st gained. The plant1s main characteristics are listed in Table I,

TABLEI

H/N Taramet er Unit Value

1 2 3 4

Reaoto1. r thermal power MW 1060 Therma. 2 l power- te supplie e th o dt chnological production JIW 440 3» Thermal power aupplied for steam generation MW 635 4. Core - pebble-bed 5. Puel element - ball-type

6. Puel element size Plffl 60 7. Core power density llW/m3 6.9 8. Core dimensions: diameter m 6.4 height m 4.5 Reacto. 9 r coolant helium 10. Pressure MPa 5.0 11. Temperature: at core outlet K 623 at core inlet K 1223 Numbe. 12 heaf ro t exchange loops pcs 4 13. Powe coolanr rfo t pumping MW 15 14. Intemediate circuit coolant - helium 15* Intermediate circuit parameters: pressure MPa 5.5 temperature at IH2 inlet K 623 at IH2 outlet K 1173. 16. Electrical power MW 265 17- Steam pressure MPa 17.0 18. Steam temperature K 813.

143 The pebble-bed core with, spherical fuel element d graphitsan e reflectors is installed in the center of the prestressed concrete reactor vessel and the heat removal equipment (IHE and steam generator (SG) is located in the side cavities of the reactor vessel /3/bypase .Th E cavitielowee IH th s e rvalven i th par se f o tsar mailowe e ane cavitiesG blowers S th dth nga e r n i th par s- f .o t The reactor unit layou presentes i t Fig.2n i d .

FIG. 2. VG-400 reactor unit layout.

1. Reactor vessel 7. Control rod (side) Cor. 2 e 8. Control rod (central) 3. Gas blower 9. Load tubes 4. Steam generator 10.Bypass valve 5« High-temperature intermediate heat exchanger 11.Unload tube 6. Ionisation chamber

144 The main building is a complez of the main and auxiliary servi- nucleae th cef o sr power plant, includin reactoe gth r building, auxiliary building, steam turbine hall, spent fuel storage, etc. reactoe th n I r buildin alse gar o installe loading-unloae th d - ding complex, auxiliary systemprimare th d f intermediatso yan e circuits, a part of the CTP heat-exchange equipment. The auxiliary buildin adjacens gi reactoe th o t r buildin d servegan accomodar sfo - tio disassemblef no d othean G r GC equipment , d SG IRE, . The maie layouth n f o buildint plane th s show i tf g o n ni Pigs. 3 and 4/4/. 7

FIG. 3. Main building of nuclear power plant VG-400. Sectional view.

. Turbogenerato1 r 9.Bridge crane 2. Bridge crane 10.Bridge crane 3. Deaerator 11.Fuel inspection equipment Stea. 4 m generator 12.Truck 5. Reactor vessel 13«Main circulator Bridg. 6 e crane 14.Spent fuel truck Leak-tigh. 7 t protection hatch 15-Fuel unloading equipment complex 8. Fuel loading equipment complex 16.Containment

The schematic diagra reactoe d layoumth an f o tr unit have been developed taking into account different stages in designing, manufacturing and testing of the plant and CTP elements. The objective

145 FIG. 4. Main building of nuclear power plant VG-400. Plan.

2 Chemoteohnologica1- l production equipment 3 Steam turbine hall Mai4 n reactor building 5 Auxiliary building 6 Spent fuel storage

of the first stage is the reactor unit tests at the medium temperature leve witd lan h minimum powenumbee th rf r o comple x elements. For this purpose it is envisaged to operate the reactor at a helium outlet temperatur f 1023Keo , with heat transferreG S o t d for steam production and electricity generation. At the second e reactostagth e cord th erean element- s de wil e testee lb th t a d signed temperatures whe core nth e outlet temperatur s increasei e p u d G outleS te o th t1223 temperaturd an K s kep i et 1023 a t addiny b K g some cold coolant bypassed from the gas blower head line to the SG inlet /5/. Tests at the first and second stages are performed with the mastered steam turbine. At the third stage the intermediate circuit will be put into operation for tests of the CTP equipment and its interaction with the reactor. This three-stage approach permits some additional time to be saved for development of high-temperature materials elementsP desigE CT ,IH d nschedule an .Th bringinf eo g temperature e th cor th o et e 1223K with hea ts show i remova G nS o t l in Pig.5.

146 0 4 8 12 16 20 24 28

FIG. 5. Primary circuit parameters when controlling by mean f by-paso s s valve.

- flow rate of four gas blowers M6g> - gas blower speed Pas - temperature difference in the core T^ - temperature at the core outlet Tj - temperature at the steam generator inlet a3 - flow rate through the core

When the valve is opened, the flow rate through the core redu

ces: G r =Gy ~ GI,V Bttcl the core outlet temperature rises

reachin 0-33« g v b 1223G core t Ka . The analysi somf so e operation regimereactoe th f so r unit with bypass showed that apart from realizatio stage-tese th f no t program the bypass scheme reveals some positive functional possibilities in the main operation regimes as well. The VG-40e safetth f 0yo plan ensures i t reliabla y b d e monitorin controd reactoe gan th f lo r unde l operatioral n conditions, precantions preventing failure reactoe th f so r element d equipsan - ment in design accidents and limitation of magnitudes of possible losses of leak-tightness in the primary circuit.

147 The safe operation of the plant is also ensured "by the inherent feature high-temperature th f so e reactor itself, which are: use of chemically inert helium having no phase transitions as the coolan d graphit tan core th e s structuraea l material, integrated arrangement of the equipment in the prestrested concrete reactor vessel and some others. One of the important conditions is creation of a reliable system for heat removal from the reactor• In accordance with the current specifications the emergency core cooling system (ECCS) must consist of several independent cooling channels and each such channel must ensure heat removal froreactoe mth r even in the case of failure of any channel, independent of the initial accident event. A schematic diagram of ECCS is shown in Pig.6. The system must ensure heat remova n losl i outsidf o s e power supply sources, loss of feed water, shutdown of the turbogenerator and heat consumers as well as in the case of the primary circuit depressurization. The coolant syste s foumha r loop mair sfo n reactor hea-é removal, which permits a part of the normal equipment to be used for ECCS, there- forlattee th e r consist independenf so t channels connected wite hth main heat removal loops capabilite .Th eacf o y h SCCS channes li sufficien residuae th r tfo l removee above b hea th o t en i dmisope - rations of the plant, except for the accident of depressurization primare th n i y circuit, when simultaneous operatio channelo tw f no s is needed. It is the equipment of the heat removal loops and not the core that imposes limitations on the increase in the reactor coolant temperature, therefore each 3CCS channel has two subsystems: that of immediate control and of durable cooldown. The first subsystem serve hear sfo t remova initiae th n - li lpe accidentae th rio f do l situatio d duratio nan s operatio it f no n (3-5 min) depend time startu f th o emergence n th so f po y diese- lge nerato d durablran e cooldown subsystem e secon.Th d subsystem repre-

148 FIG. 6 Schemati. c diagra VG-40f o m 0 reactor emergency cooling system.

1. Reactor 2. High-temperature intermediate heat exchanger 3. Steam generator 4. Gas blower of the primary circuit 5. Condenser 6. Condensate pump 7. Hydroaccuiuulator.

sent sclosea d circulation loop designe removar reace dfo th -f lo tor heat to a safe level that practically corresponds to the amount of losses into the environment.

149 REFERENCES

1. V.N.Grebennik "Status of Development of Gas-Cooled reactors in the USSR". Paper presented to VII International Conference on Gas-Cooled Reactors", PRG, Dortmund, 17-18 Sept., 1985. 2. F.M.Mitenkov, et al."Features of the principal scheme and construction of the pilot high-temperature gas reactor". Voprocy atomnoi nauk tekhnikiii . Ser.Atomnovodorodnaya energetika, 1977 2(3)K , , p.15-1 Russian)n 8(i . N.N.Ponomarev-Stepno. 3 . "Designal t ie reinforcef so d concrete vessels for new national HTGR", ibid., 1984, IT 2 (18), p.30-32 (in Russian). 4« M.I.Zavadskii, N.V.Sukhoruchenkov, S.V.Shishkin . "Teal -t ,e chnoeconomical prerequisites for application of the high-temperature gas-cooled reactors in power-intensive fields of industry" The Japanese-Soviet seminar on many-purpose application of the reactors, Tokyo, 25-27 January, 1983. 5. F.M.Mitenkov, et al. "The nuclear energotechnological plant VG -400-substantiation of scheme solutions". Voprocy atomnoi nauk tekhnikiii . Ser. Atomno-vodorodnaya energetic tekhnoi a - logiya, 1985, N 2, p.59-66 (in Russian).

150 CURRENT STATU RESEARCF SO DEVELOPMEND HAN T OF HTGR IN JAPAN

T. HAYASffl Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, Japan

Abstract

In Japan, the very high-temperature reactor ( VHTR ) for a nuclear process heat applicatio s beeha nn develope y Japab d n Atomic Energy Research Institute (JAERI) since 1969. Looking back upon the activities, there are considerably accumulated technologies enoug o construct h e experimentath t l VHTR. Recently, however, economic structure in Japan has been change w growtdlo wite hth h rat e nation'f economo eth d an ys energy demands have been relieve o rapidls d y thae needth r t fo s nuclear process heat application have been declined. This trend alss beeha o n accelarate y introducinb d g firmle th y energy-saving technologie n industri s d witan y h increasinf o g the electric power generation by light water reactors. In this situation, the Special Committee of Japan Atomic Energy Commission (JAEC) has recognized newly the significance e VHToth f R development froe viewpointh m f lono t g futurd an e has recommended the construction of a high-temperature engineering test reactor (HTTR o upgrad)t e currenth e t VHTR technologies. JAER s starteIha w deploymenne d a researce HTT s th a R f o t h facility to prepare for the realization of nuclear process heat applicatio e 21sth tn i centuryn . This paper deals with technical background and the current design concept of the HTTR.

1. Introduction

In Japan, since 1969 researc d developmenan h r verfo t y high temperature reactors (VHTRs) has been carried out for about seventeen years.

151 In 1970s, Japan Atomic Energy Research Institute (JAERI) conducte desige th d n stud d developmenan y e experimentath f o t l VHTR which had the output power of SOMWt with the reactor coolant outlet temperature of 1000°C in order to demonstrate as a heat sourc f reducineo productios ga g r irod steelmakfo n an n - ing industries. In parallel with these activities in JAERI, the development on utilization system of the nuclear heat was starte n 197y Engineerini db 3 g Research Associatio f Nucleao n r Steelmakin g projecg bi (ERANS e n Ministri th t s a ) f o y International Trade and Industry (MITI). The results accomplishe n thii d s project were transfere o JAERt d I aftee th r terminatio e projecth f n 1980o ni t , whee phase-th n I program was completed.

In 1980s, economic fluctuation in Japan has been induced growtlow witeconomy of henergy-savinthe hthe and , g technologies have been introduced thoroughly into industrye th s A . resul f theso t e situations, nation's energy demands have been relieve o rapidls d y thae need th r tnuclea fo s r process heat applicatio n industri n y have been declined.

n AprilI , 198 e Committeth 6 n Nucleao e r Process Heat Utilizatio e Japath nf o nAtomi c Industrial Forum (JAIF) issued e reporth n perspectiveo t e nucleath f o s r heat application i n industry, in which the Committee emphasized the necessity of continuatio develoo t n e indigenouth p s technolog f VHTRo y n i s preparing for the coming century.

In connection with the view of industry, Japan Atomic Energy Commission (JAEC) started the Special Committee on HTGR Research and Development Program in April, 1986 to investigate clearly a significance of the VHTR development in Japan.

After serious discussion for four months, the Special Committee submitte a dinteri m e reporJAEC th n thi I o .t t s repor Committee th t e recognize e definitth d e significancf o e the.VHTR development in Japan and recommended that it was appropriat construco t e a higt h temperature engineering test reactor (HTTR) as a research facility to replace the experimental VHTR schemed by JAERI.

152 Base n thio d s conclusion t ,abou se e JAERcon th s -t ha I struction program of HTTR to use it for the purpose of not only upgradin e currenth g t technologie t alssbu o basic researchen i s cooperation with universities and national institutes. JAERI is now about to solidify a design concept of HTTR by qualifying the current technologies develope r seventeefo d n years.

2. Technical Background

Looking back upon the development activities in Japan, main achievement e outlinear s s followsa d :

2.1 Research and Development in JAERI

e beginningIth n e effortth , s were concentratee th n o d design of the experimantal VHTR supplying the thermal output of 50MW with the reactor coolant outlet temperature at 1000*C. The functions of the experimental VHTR were as follows:

1 ) To demonstrate the technologies for nuclear process heat applications, 2) To confirm the safety of VHTR plants, 3) To develop the technologies for proto-type VHTRs.

However, following the recommendation of the Long-Term Program for Nuclear Energy Development revised by JAEC in 1982, the reactor coolant outlet temperature was changed from 1 000°C to 950' g e projec Cirod steelmakinbi th sinc an ne r th fo et g application was terminated in 1980.

Afterwards e desig,th n efforts have been continuee th d an d detail design of the experimental VHTR has been completed in 1985.

In parallel wite desigth h n works, researc d developan h - menbees ha tn carrien fuelo d graphit t an s ou d e materials, heat resistant alloys, components and structures, reactor engineering, irradiation testa hig t a hs temperature d evean , n thermochemical hydrogen production.

153 e irradiatio th o servT r efo n e coatetestth f so d particle fuel d othean s r material e temperaturth t a so 100Q1C t e p u eth , in-pile heliu s looga mp OGL-1, installe Japae th t na d Material Testing Reactor t (JMTR)intpu o s operatio,wa 1976n i n . After numerous irradiatio h irradiatioOt 1 n e e OGL-workth th , 1n i sn test of the fuel compacts was finished in May, 1986.

For the purpose of clarifying the behaviors of the graphite materials in the core environment, extensive studies have been in progress on chemical and physical properties as well as mechanical properties. The results through these studies have been presented in the IAEA Specialist Meeting. e heath tr resistanfo s A t alloy r structurafo s l applications e long-ter,th m performance tests including corrosion, carburization, creep, stress-rupture and thermal aging characteristic e Hastelloy-Xth n so R have been conductede th r Fo . development of second generation superalloys, the various tests to assess the feasibility of nuclear reactor applications on the five kinds of Ni-Cr-W alloys with different Cr/W ratio have been examined for exploring the optimum ratio for balancing between creep strengt d corrosioan h n resistivity.

Concerning the development of components and structures at a high temperature e Heliuth , m Engineering Demonstration Loop (HENDEL), which consist e Motheth f o sr (M), d Adaptean ) (A r Test (T) sections, was put into operation partially in 1982.

sectioA e HENDELM+ th e f no Th , which provide e heliuth s m gas at a maximum temperature of 1 000*C with a maximum flow of 4kg/s, has been operated for more than 6,300 hours so far. In the fuel stack test section (T^) e thermo-hydraulith , c testf o s the mock-up fued controan l l rod n reactoi s r operating conditions were finishe n 1985i d .

The constructio tese th t f sectioo n r in-corfo T n e struc-

ture tests was completed2 in June, 1986, and test operation for abou 3 hour4 a temperaturt t a s f 1050to e s successfull:wa y proceeded.

154 The test items in the T2 test section are: 1) Coolant leakage through the graphite block structures, 2) Performance of the core support structure, Coolan) 3 t mixing behavio e high-temperaturth t a r e plenum.

Other test sections for the high-temperature components of the heat removal system such as an intermediate heat exchanger an a stead m e constructegeneratob o t e nea e th rar rn i dfuture .

In the area of reactor engineering, critical experiments % enriche20 e oth nd uranium graphite-moderated cores simulated o VHTt R have bee criticae nth carrien i lt assemblou d y (VHTRC) which was reconstructed from the experimental facility, Semi- Homogeneous Experiment (SHE).

For the reactor instrumentation, the high-temperature fission counter and gamma-compensated ionization chambers have been develope withstano t d a temperatur t a d o SOOtt p Xeu

In addition, the miscellaneous developments have been e thoriuth carrien o m t fuelou d , fuel reprocessing, coaxial double pipe, plate out of fission products and thermochemical hydrogen production, etc.

2 Researc2. d Developmenan h t Activit ERANn i y S

The purpose of the project was to establish the fundamental engineering technologies requireconstructioe th r fo d f no the nuclear steelmaking pilot plant, which would be connected wite experimentath h l VHTR being develope y JAERb d o t I demonstrat firse th e t applicatio nucleaf no r heat.

Main item d obtainean s d results wer s followsea : 1) High temperature heat exchanging system An intermediate heat exchanger (1.5MWt a hig d han ) temperature helium test loop were constructed and operated successfully from 1978 to 1980 by Ishikawajima-Harima Heavy Industries.

155 2) Heat resistant superalloy Several kind f alloo s y were tested with respeco t t e requireth d properties o alloystw d ,an , KSd Nan SSS113MA, were selecte e candidat th e heas th a d tr fo e transfer tube of the intermediate heat exchanger. These were concerne y Sumitomb d o Metal Industries, Hitachi Metals Industries, and Kobe Steel Industries.

3) High temperature heat insulating materials Development works were conducte n aluminoo d - silicate fibrous materials and fused silica type refrectories by Isolite Babcock Refractories Co. and Toshiba Ceramics Industries.

4) Reducin s makinga g g unit Research and development was carried out on making e reducin th y steab s m ga greformin f ligho g t hydrocarbons and gasification of the pitch, which were obtaine y steab d m cracking. Steam reforming test plant and pitch gasification test plant were constructed an d operated satisfactorily by Chiyoda Chemical Engineering & Construction Co. Ltd.

In addition, the conceptual design of the nuclear steelmaking pilot plants, FM-50 syste d UC-5man 0 system s carriewa , d t sufficientlou s prearrangea y e Grouth f y Systeo pb d m Analysis.

These results achieve e prograth n i dm were transfered from ERAN o JAERt S I afte e terminatioth r e prograth f n 1980o ni m .

2.3 Development in Industry

Most of nuclear industries in Japan have been concerned the development of VHTR technologies in cooperation with JAERI and/or ERANS.

t MitsubishA i Heavy Industrie high-temperatura s e helium gas loop was installed in 1977 to test the perfomance of com-

156 ponents, behaviors of metallic and graphite materials at a t\ temperatur o 1000ta pressurt d p o u e49kg/crrrG;t an p u e .

n 1978I , a high-temperatur e helium test loop (EM-1.5s )wa constructe t Ishikawajima-Harima d a Heavy Industrie o develot s p He/He intermediate heat exchangers and steam generators under the contract with ERANS.

n 1979I a high-temperatur, d high-pressuran e e helius ga m loop (KH-200) was instralled at Kawasaki Heavy Industries to tese dynamith t c performanc f heliuo e m flow, endurancf o e components and verification of dynamic analysis calculation codes.

Babcock-Hitachi Industries also bult a high-temperature helium gas Loop (BHK) in 1984 to develop He/He heat exchangers connected with side-by-side piping, steam generators and steam reformers.

e datTh a attaine n thesi d e development e use- ar sin d dividually for design of the reactor systems conducted under the contracts between JAER d industriesIan . r fuel fo d graphit s an sA e materials, Nuclear Fuel Industries (NFI) fabricated the mass production system of the coated particle fuel in 1973 and the nuclear grade graphites d IG-11IG-an 11 0 have been produce y Toyb d o Carbo . sincCo n e 1975. These products are provided for the various experiments in OGL-1, VHTRC, HENDEL, etc.

3. Perspective n VHTso R Development

Special committee of JAEC issued the Interim Report in August, 1986. Outline Reporth s f summarisei o te s followsa d :

157 3.1 Significanc f Researceo d Developmenhan n VHTRto Japan si n

In general essencialls i t i , y importan r promotiofo t f o n nuclear energy development to improve the economy and to expand an application of nuclear energy.

Nowaday n Japani s , electric power generatio y lighb n t water reactors is established well so that the total installed capacity is 24,686MWe with 34 units, which the gross generated outpu totatn i take% le shar 26 electricitsth f eo n 203i 0d an y the share of nuclear power supply is projected to be 58%.

Looking at a future situation, if the thermal efficiency of plant is more increased by producing high temperature nuclear heat, it can be expected to improve the economy and to expand applications of nuclear energy, which will be resulted in improvemen r livinou f to g environment.

e mose VHTth Th s tconsidere i Rf promisin o e on s a dg nuclear reactor o solvt s e these probleme sth becauss ha t i e characteristics of highly-inherent safety as well as high- temperature nuclear usefue heab o broar tt lfo d range.

n JapanI , however e definitth , e tim o introduct e e th e nuclear process heat in industry is so uncertain at present that the steady promotion of the VHTR research and development is very significant from the viewpoint of long future.

In addition, Japan is now facing the time to encourage creative researche r innovatiosfo sciencn ni d technologean o ys that national research institutes and universities as well as industries should promote basic researche producinr sfo g seeds w technologiesone f .

Along these trends expectes i t ,i d tha e researcth t d an h developmen VHTRn o t s will giv remarkabla e e effec n technicato l innovation in the future.

158 2 Futur3. e Perspectives

Based on the significance of research and development on e revieth VHTR f technologied wo an s s attaine o nows t i p t ,u i d recommended that a high-temperature engineering test reactor (HTTR) should be constructed in JAERI as early as possible and more efforts should be devoted to upgrade a VHTR technology on the basis of the accumulated technologies and human resources so far in Japan.

Main subject upgradee b s follows a o t se ar d : o 1establisT ) h full e indigeneouth y s technolog f VHTRso y , 2) To demonstrate the highly-inherent safety of VHTRs such s passiva e safety, 3) To improve the core performance such as higher coolant temperature, fuel burnup, power density, etc., 4) To improve the thermal efficiency by use of high- temperature nuclear heat, o serv T r irradiatio) fo 5 e n tests.

Moreover, it is instructed that the operation of HTTR should be carried out in cooperation with national institutes and universities which have the requirements to use the HTTR for basic researches.

Consequently, the core structure of HTTR is to be a prismatic block type that is suitable for conducting the various irradiation in the core and then the thermal output is to be 30M t leasa Wr settinfo t e irradiatioth g n regioncoree th .n i s w PrograNe . JAERn 4 i m I

n accordancI e wite Interith h m Report, JAER shiftes Iha d the researc d developmenhan t program froe previouth m s planning of experimenta w deploymenne a e e HTTl s th th VHTa R o f t o tR research facility.

Forcusin e constructioth n o g f HTTRs reconsidereo ni t ,i d modifo t previoue th y s design concepe experimentath f o t l VHTR.

159 Implementin a desigg f HTTRo n , there shoul e takeb d n account of on the items as follows:

1) To accomplish upgrading technologies of VHTRs, 2) To serve for the various irradiation tests, o prov3T )e usefullnesth e e nucleath f o s r heat application in future.

Therefore, regarding a current concept of HTTR, the basic specifications are shown in Table I, in which the reactor output power is 30MWt with the helium outlet temperature up to

TABLE I BASIC SPECIFICATIONS OF HTTR

MAIN ITEM DESIGN CONDITION

1 REACTOR THERMAL OUTPUT 30MW

2 REACTOR OUTLET COOLANT 850t~950

3 REACTOR INLET COOLANT 3951C TEMPERATURE (AVERAGE)

4 FUEL UO2

5 FUEL ELEMENT TYPE PRISMATIC BLOCK

6 DIRECTION OF COOLANT FLOW DOWNWARD W THROUGLO -F H THE CORE

7 PRESSURG VESSEL STEEL

8 NUMBER OF PRIMARY COOLANT 1 LOOP CIRCUIT

9 HEAT TRANSMISSION IHX AND PWC (PARALLEL LOADED)

10 PRIMARY COOLANT PRESSURE 40kg/cm2G

11 SECONDARY COOLANT He PRESSURE IS HIGHER PRESSURE THAN PRIMARY COOLANT PRESSURE

12 COMPONENTE TH N I S PWC, CIRCULATOR. ETC. SECONDARY CIRCUIT

13 CONTAINMENT TYPE STEEL CONTAINMENT

14 PLANT SERVICE LIFE 20 YEARS

160 950"C. Fig show1 . s coolant diagra e systemth f s showo mA . n i n Fig. 2, the core of HTTR is composed of the hexagonal fuel block controd san l blocks. Fig show3 . e verticath s l vief o w the core contained in the pressure vessel. The reason why we choos e cor th f eprismatio e c block typ s basei e n o d Heat izatioI i t U n System r——i He Circulator He Circulator

30 MWt REACTOR 1C 10 MWt

PWC '— — AHX Vessel IHX Cooling 'System r MWt O 10 MWt 0 He Circulator /

t MW 0 3

PWC 9501C- 395 1C He Circulator

Fig. 1 Coolant Flow Diagram of HTTR

Nd Name 1 Fuel block 2 Control block 3 Replaceable reflector 4 Permanent reflector 5 Side shield 6 Reactor pressure vessel

Fig. 2 Cross Section of HTTR Core

161 Na Name 1 Fuel region 2 Removable reflector 3 Permanent reflector 4 Hot plenum block S Cote restraint mechanics 6 Reactor pressure vessel 7 Diagrid 8 Outlet pipe g Stand pipe 10 Flow control unit

Fig 3 . Vertical Vie f HTTwo R availability and flexibility for various irradiation tests. As for the circumstance of neutron irradiation in the reactor, there are three regions, namely the regions of core, replaceable reflector permanend an s t reflector s showa s n Tabli n , II e in which it can be estimated that the maximum flux of thermal neutron is around 8 x 10l3n/cm^-s and the maximum helium temperature is about 1100°Cat the assumed core.

162 TABLE n IRRADIATION CHARACTERISTICS OF HTTR CORE REGION

Irradiation Core Region Replaceable Reflector Permanent Reflector Region Fued ColumRo l m Contro Columd Ro l m Repl. Ref. Columm Control Rod Columm Columm Irradiation Condition Hexaganal Block Cylindrical Space Hexagonal Black Cylindrical Space Cylindrical Space Style Width Across the Width Across the Size Flats : 360 mm m Diametem 0 11 : r m m 0 Flat36 : s Diameter '• 110mm Diameter : 100mm Height : 722.5 mm Height : 700 mm Height : 722.5mm Height : 700mm Height 13,000 mm Nimber of Specimens (Max) 24 7 48 12 12 Max. Neutron Flux 2 2 u z FastOO-lMeV) ~2xlO'3n/cm 's H —— > ~lxlO»n/cm 'S ~3xiO n/cm -s Thermal (< 0.4 eV) ~8xi013n/cm2-s ~7xlO'3n/cm2'S ~4Xl013n/cm*-s Temperature at e Regioth n 4 00 - 1,1 00 1 400~1.000"C 400~900"C 400~900t 400-600X; Instrumentation Installee th n i d Measurement f Temperatureo s , Thermal Neutron Flux. Neutron Moni tord Fuean , l Failed Detector Core Examplf o e Block Type Fuel, Fuel Compact. Ceramics. Irradiation Graph. te i -H. . _ Metal Specimens Pebble Type Fuel Metal, Ceramics

Concerning the construction schedule of HTTR, as shown in Table III, licensing review by Nuclear Safety Commission is to be conducted durin gconstructioe FY198th d e HTTan 8 th s i R f no projected from FY1989 to FY1994.

TABLE CONSTRUCTION SCHEDULE OF HTTR

FISCAL YEAR '86 '87 '88 '89 '90 '91 •92 '93 '94

Modified Design l—— — —— 1

Final Design i— — \

Geological Survey 1—— — —— l

Licensing —— 1

Construction l— — ————— 1

Fuel Fabrication

Conclusio. 5 n

Based on the recommendation of the Special Committee, JAEC recognize e significancth d e VHTth Rf o edevelopmen t froe th m viewpoint of long future in Japan.

163 n predicAtca presene on t o definiteln t a economiy c stability under the low price of oil in the 21st century. Althoug assumes i t i hJapan i d n that o therexpectation s i e o nt make a practical use of the nuclear process heat within this century e mos th e VHT tconsideres ,th f i R o promissin e on s a d g reactors since it has highly-inherent safety and potentiality expano t nucleae markee th th d f to r process heat application. These e VHTRsense th e bein n ar so s g emphasized because much more it is very important to realize an intrinsically safe reactor now that many peoples have experienced two severe accidents such as TMI-2 in 1979 and Chernobyl in last April. e RussiaTh n "Chernobyl" meansu knowyo s ,,a "Wormwood" whics i h indicated in the Revelation to John of the New Testament, namely that "The name of the star is Wormwood. A third of the waters became wormwood e water th d man n - dier ,an me be ,yfo d cause it was made bitter." The prediction has come true.

Nowadays it seems that we are about to get into new age that the small is beautiful since small and medium reactors are strongly expectet onlno yr electrifo d c power generatiot bu n also process heat application in industry.

Therefore n Japani , t shouli , e emphasizeb d d thae th t importanc e VHTth R f o inveso edevelopment muc o s w ho h-t no s i t tigat econome nucleae th eth f o yr process heat applicatios a n how to establish the indigeneous technologies as early as possible in order to cooperate with the advanced countries, so that it is actively expected for the HTTR to play an important part at the frontier of the VHTR development toward the 21st century.

164 PLANT DESIG HIGA F NHO TEMPERATURE ENGINEERING TEST REACTOR IN JAERI

. BABO A Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken, Japan

Abstract

Special Committe n HTGo e R Researc d Developmenan h t Program submitten a d interim report after serious discussion which recommended that a high temperature engineering test reactor (HTT thin i R s report) shoul e builb d o t t make possible irradiation tests not only for HTGR fuel but also for various materials in high temperature. The required functions to HTTR and its basic design condition are described together with selecting progress A preliminar. y desig a hig f o hn temperature testing reactor, which makes possible HTGR fuel irradiation test d demonstratioan s n test f HTGso R inherent safety s beee ha ,n th dond an e results are also explained.

1. Introduction

In March 1986, Special Committee on HTGR Research and Development Program, which was set up by Japan Atomic Energy Commission (JAEC), started e e revieresultth th n o f wpasto s activitie n VHTo s R developmen d energan t y situation in Japan. The committee has submitted an interim report to JAEC in August. The report recommends that it is appropriate to construct a high temperature gas-cooled testing research reacto researca s a r h facility. Accordin e recommendatioth o t g n JAER s starteha I e desig a higth d f o hn temperature engineering test reactor (HTTR e Multi-purposn placth )i f o e e Very High Temperature Experimental Reactor (VHTR-Ex), which JAERs ha I developed since 1969.

2.Functions of HTTR

Major functions which were required to HTTR in the interim report of the Special Committe s followsa e ar e :

Category A. For upgrading present HTGR technology, a. Irradiation tes f HTGto R fuegraphited an l , b. Demonstration tes f HTGo t R inherent safety, c. Demonstration test of high temperature heat utilization system,

165 Categorr variouFo . B ys progressive fundamental researche n i higs h temperature, a. Irradiation test in the field of material science and technology, b. Irradiation test in the field of fusion technology, c. Irradiation test in the field of radio-chemistry.

The former functions (A. above) have been attached to the VHTR-Ex. The latter functions (B. above) are newly added reflecting the needs in Oapan. These haven't been taken into account for ithe design of the VHTR-Ex, therefore, GAERI has to change its design to make possible these irradiation t clea no t wha s ye ri t t testskini f irradiatios o dA . n test e possiblar s n i e what condition, OAERI is going to decide irradiation condition, capacity and limit in HTTR in detail after feasibility study. The irradiation tests which are subjected to the feasibility study at present are listed up below:

. Irradiatioa f higo n h temperature alloys, carbo d ceramian n c materials, b. Irradiation of fusion reactor blanket, c. Irradiation tes f higto h temperature wadiation resistant detectors, d. Demonstration test of contenuous Tritium production and recovery technology, e. Production of radio-isotopes, f. Various materials irradiation tests for radiochemical hydrogen production, thermal splittin f taro g , etc..

. Basi3 c Design Conditions

In June 1986, JAERI has finished a preliminary design of a high temperature testing reactor which has the functions of category A. As mentioned above, this design should be chagned in detail after feasibility study in orde o fulfilt r e functionth l f categoro s . B y However, as it is expected that the basic design conditions will not be changed, here is written the selecting progress of design condition.

) Therma(1 l Power Reactor core performance such as core life, shut down margin including rod-stuck margi maximud an n m fuel temperatur examines ewa r fo d three cases, 50Mw, 30Md lOM f an wo thermaw l power. Irradiation test capability of HTGR fuel element and graphite block was also examined.

166 The target for core life was set up to 660 effective full power days and e maximu1300°th r fo Cm fuel temperatur operation i e n conditio f coro n e average power density and core outlet coolant temperature at 2.5 w/cc and 850°C respectively. The reactor core of lOMw did not fulfill the target values of core life and reactor shut down margin. The maximum fuel temperature was lower than 1300°C, but higher than that of 30Mw core. The core of 30Mw didn't meet the irradiation conditions of VHTR fuel, but it became clear that 30Mw core can give better irradiation condition than 50Mw core.

) Reacto(2 r Outlet Coolant Temperature e outleTh t temperatur s surveyewa e d froe vieth m w poin f definino t g the lowest temperature which can be used for hydrogen production as a case of nuclear process heat direct application. For various hydrogen production processes with thermochemica1 reaction e highesth , t temperatures of each processes were compared. The UT-3 process, which is developed in Tokyo University in Japan, was found to have the lowest reaction temperatur e highesf 730°o th e t Ca t e processpointh n i t . Reactor outlet coolant temperature e 850°b , o thereforeCt t se s wa , taking into account primary and secondary coolant temperature difference at an intermediate heat exchanger (IHX) and temperature drop at piping.

(3) Configuration of Reactor Cooling System e configuratioTh f reactoo n r cooling syste s examinewa m d froe th m aspects of safety, required function, plant control abi 1 ity and construction cost. There were mainly two possibilities in the selection of core cooling method in case of failures of main cooling system. One of them was that the reactor vessel cooling system (VCS) cools reactor vessel an casel d al corf mai o sn i en cooling system failures wa e othe Th e . on r that a forced direct core cooling loop in safety class cools the reactor core when flow path is sound and VCS cools the core in other case. After core cooling analyses e latteth ,s selectewa r o prevent d t damage f controo l rod. Numbers of main cooling loop, safety class forced cooling loop and 2 respectivel d an 1 , 1 o S minimiz wert yt VC se e e construction cost. There were many possibilities wit n IHXa h . Heat exchange capacity e importanth f o e ton discussios wa n points. IHX f 25Mo s w have been

167 designe r lonfo d g tim n VHTR-Ei e n ordei x o havo t rsymmetr tw e y main cooling loops. An IHX of 1 OMw was selected as a variation for it considering a scale of process heat plant and heat loss in the piping. Another possibility with IHX was construction schedule. One of the options was to build it from the first of reactor construction and another was to build after operation of the reactor for certain duration in order to decrease initial construction cost. Cases which had no IHX r wholfo e reactor life were also discussed. Final decisio r thifo n s discussion wil e donb l e after comparison between whole construction cost and budget.

(4) Containment Type

Licensability of exclusion the rapid depressurization accident in primary cooling syste s examinedwa m . e materialTh s use reacton i d r coolant pressure boundar e welar y l known their propertie d welan s l experience n nucleai d d chemicaan r l industries. However the possibility of ductile ustable rupture may be very low, JAER Ie conclusio th cam o t e n tha might.bt i t e very difficult o provt e probabilitth e o establist yd enougan w hlo h leak before break criteria becaus f loc o ef baco k k data. After irradiation dose analyses for public in case of rapid depressurization accident, containment vesse s selectedwa l .

) Selecte(5 d Design Conditions

Typical cases with combination f desigo s n parameters described above are shown in Table I. Construction cost for each case was roughly estimate comparedd an d e numberTh . s shoe ordeth w r from lower cost. Finally, the case 2 was selected and other important design conditions were decide s showa d Table-IIn i n .

3. Preliminary Design Result

(1) Cooling System

The cooling syste composes i m maia f no d cooling system (MCS)n a , auxiliary cooling syste non-safets mi o VCSs(ACSS tw MC d . )an y class and ACS and VCSs are safety class as explained above.

168 Tabl I e Typical Combinatio f Desigo n n Parameters

Case 1 2 3 4 5 Thermal Power (Mw) SO 30 30 30 10 Outlet Temp. (°C) 950 850 850 850 850 (Mwt) 10 10 10 IHX Set up after with after Schedule operation reactor operation

^BVCS RVCS firv / EÄE JHX-SG Wf-HX AHT XT Cooling System LjMpvvc @ A PWC PWC ® PWC PWC

Construction Cost 5 3 4 2 1 (number from low)

Table II Basic Design Conditions of High Temperature Testing Reactor

Item Basic condition

1 Reactor thermal output w M 0 3 2 Reactor outlet coolant 850°C, 950°C temperature (average) 3 Reactor inlet coolant 395°C temperature (average) 4 Fuel Low enrichedU ,UC 5 Fuel element type Prismatic block 6 Directio f coolanno t flow Downward- flow through the core 1 Pressure vessel Steel 8 Numbe maif o r n cooling 1 loop 9 Heat transmission C (parallePW d IHan X l loaded) 10 Primary coolant pressure 4.02 MPa 11 Secondary coolant prssure He pressure is higher than primary coolant pressure 12 Containment type Steel containment 13 Plant life 20 years

169 o heatw ts transfeha S MC r paths from reactoe r th cor o L e environment air. One of them is from reactor core to air cooler via primary pressurized water coole rs fro i e othe(PPWC mth e d on r )an reactor core to air cooler via IHX and secondary pressurized water cooler coole r (SPWC) ai r bot e fo r hTh .path d SPW commons an i s C X IH . have samth ee heat transfer capacit f lOMwo y o .mode tw f PPWs o s ha C operation whic e 30Md 20Mwar han w d .secondar an WheX IH n y helium loop are operated, PPWC is operated at 20Mw in the same temperatures at primary helium inlet and outlet of PPWC as operated at 30Mw. In order o makt e possible thes o modetw e s operation o primar,tw PPWs ha Cy helium outlet nozzles eac f whico h uses r 20Mi h fo r d30Mo w w operation. Main reasons for two modes operation are,

(a) to operate reactor at 30Mw even in case the IHX is not built at the same time with reactor, t simplege d o clearet an r ) (b r condition n safeti s y experimenty b s operating PPWC at 30Mw, in which thermal transient might be much severer than normal operation.

Reactor inlet temperature is controled by changing inlet water temperature at PPWC and secondary helium flow rate for IHX. All of the gas circulator d watean s r pump e operatear s d with commercial electricity without emergency back up. ACS consists of an auxiliary heat exchanger (AHX), auxiliary gas circulator independens r cooleri s ai S (AGCsn AC a . d )an t system from MCS and is normally in stand-by condition. ACS is operated only when MCS has some failures or the reactor is shut down and MCS is in overhaul. All of the active components such as AGC and water pump are redundant and have emergency back up power supply. A flow diagram of cooling system is shown in Fig. 1, and a heat and mas st paralle a balanc S MC lf o eoperatio d PPW ns an show i witCX n i nIH h Fig. 2. The two VCSs have the same heat removal capacity of 100% to cool reactor vessel and core in emergency. Both VCSs are normally operated t 100a % flow rat n ordei e r cool biological shield aroun e reactoth d r vessel, therefore no start-up motion is required at the accident.

170 Containment Vessel

Water Pump

. i . Coolinno g System

SE1C SGC T 395 P 4.04 G 14.6

R/V

i 395 P 4.04 29.2

T 140 T TempC ° ; . P 3.53 P o PressMP ; . G 412 G Flowh / Rar t ; e

FIG.2. Masd Heaan s t Balanc) (I e

171 (2) Reactor Building e reactoTh r building contain a containmens t vessel, sub-systemr fo s cooling systems, ventilatio r conditioninai d an n g systems, reactor control room, fuel handling and storage facility, etc. in order to simplify asseismic design e reactoTh . t almosa r e t vesseth pu t s i l e centrreactoth f o er buildin s mana g y LWR Japans i s o thas e ,centr th t e f gravito f reactoo y r builin s e almosreactoi gth n i t r vessea d an l vertical e reactomotioth f o n r core cause y rockinb d g e motioth f o n building during earthquak s minimizedi e s showA e .n Figth i n , 3 . containmen g nozzltbi vessea e s s veri abovha l yt i smale d reactoan l r vessel. The cover of the nozzle is removed when fuels are exchanged. The big room over the operating floor is so designed that pressure insid s maintainei e d slightly lower than outside.

(3) Reactor Core and Vessel Reactor fuel is a prismatic block type so as to serve irradiation spac r HTGfo e R fued graphitan l e block n activA . e cor s sorroundei e y b d

Replaceable .Spent fuel reflected strage pool s trage

Emergency electric power supply

Commercial electric power supply

Helium purification Battery room system

:•••.•;"/ L •3180E 0

FIG . 3 .Cross-Sectio f Reactoo n r Building

172 replacable and permanent reflector blocks. The reactor pressure vessel Is made by a normalized and tempered 2 1/4 Cr-lMo steel. Reactor coolant flows inte pressurth o e vessel throug a nozzlh e attachee th t a d bottom and flows up inside of the vessel in the mean time it cools core restrant ring t flowI . s downward througcoolanl al e cord s th hi tan e guided t plenuintho a o m t flowbeloi t e corthougd ou s th we an sam th he nozzle at the bottom but inside of the co-axial double nozzle. Coolant of ACS flows almost the same path with that of MCS but through different nozzle A vertica. t dowcu ln cross sectio s showi n n Figi n . .4

Stand pipe

Permanent reflector

Fuel region

Removable reflector Core restraint mechanics

Reactor pressure vessel

Hot plenum block

Outlet pipe

Dia grid

no.4. Reactor Vesse d Coran le

173 ) Component(4 Coolinn i s g Systems The IHX is a vertical helical coil counter flow type heat exchange shows a r n i Fign . .5 Primary coolant flows shell sidd an e secondary coolant tube side. Materials for cold pressure boundary and very high temperature structures are 21/4 Cr-IMo steel and Hastelloy-XR respectively. Primary hot gas flows into IHX through inside of a co-

A Secondary He Outlet

Secondar| 4- y He Inlet

Filter

Gas Circulator

Motor

Insulator

Prinary He Outlet

Primary He Inlet

Intermediat. FIG5 . e Heat Exchanger

Heat Transfer: 10HV

174 axial double e bottonozzl th d flow t an ma eoutsidp u s f heao e t transfer tubes exchanging heat with secondary coolant in the tubes. Primary hot gas flows out through a nozzle at the top of IHX vessel side wall and pressurized by MGC and flows in again into the IHX vessel. It flows down ring space between inne d outean r r vessel d goe t througan ssou e th h bottom nozzle. The flow path is more complexed than the IHX tested in ERANS (Engineering Research Association of Nuclear Steelmaking), but it s thoughwa e failuret th safe r t fo thermara s l insulato d innean r r vessel. PWCs and AHX have very similar structure which is a vertical U tube type heat exchanger. Helium gas, which is primary coolant for PPWC and d secondarAHan X y coolan r SPWCfo t , flows outsid f heao e t transfer tube and pressurized water flows inside of the tube. Flow path of helium gas is similar to that of IHX on the same reason. The PPWC has two kinds of helium outlet nozzle adjuso t s t heat transfer are o 20M t ad 30M an ws a w explained before. (Fig) 6 . Heat Tuba

Prinare H y Outlet

Circulatos Ga r Prinary He Inlet Motor

Insulator

Tube Sheet

Vater Inlet Water Outlet.

FIG. 6. Pressurized Water Cooler (PWC) Heat Transfer: 30KV

175 Co-axial double pipe is adopted for high temperature gas transfer whic s i showh n Figi n t . aboua .7 Cols ga td 400° betweep C ga flow e nth s inne outed an r r pipes toward reacto fro s e reactomga th t r ho inle d r an t flows e insidinneth f reo pipe. Temperatur f inneo e r pip s almosi e e th t same as that of outer pipe because its inside is thermally insulated. Difference of thermal expansion betweem these pipes is absorbed by a flexibilit f inneyo r pipe. Therefor compensatoo n e bellows a r s joins i t used. This is also the reason why coolant nozzle of reactor pressure e bottomth vesset a . s i l

Biological Shield

IHX

High Temp. Helium (to IHX) High Temp. Helium Low Temp. Helium (from reactor) (from IHX) w TempLo . Helium « High Temp. Helium Ï8636 (to reactor)

Outer Pipe Inner Pipe Liner

. 7 FIGCo-axia. l Double Pipe

4. Conclusion

A preliminary design of a high temperature testing reactor has been done successfully which has the functions of category A. As described in 2. Functions of HTTR, the HTTR has to have the functions of category B which includes irradiation test f variouo s s kin f materialo d r individuafo s l purpose.

176 JAERI has to define the irradiation condition which can be given in the HTTR after feasibility study. Afterwards JAERI is going to give necessary change to the basic conditions in detail and the preliminary design written in this report.

. Acknowledgemen5 t e workth f so writte l Al n thii n s e memberreporth don s y b en wa i VHTt s R Designing Laborator f JAERo y n co-operatioi I n wite memberth h n Fuji s i Electric Company and Mitsubishi Heavy Industry. Author thanks all of them.

Reference

1) T. NAKAYAMA, H. YOSHIDA, H. FURUTANI, H. KAMEYAMA AND K. YOSHIDA. MASCOT-A Bench-scale Plant for Producing Hydrogen by the UT-3 Thermochemical Decomposition Cycle, Int. J. Hydrogen Energy, Vol. 9, . 187-190PP NO, 3 . , 1984.

177 HTR-MODULSTATUE TH F SO E PLANT DESIGN

LA. WEISBRODT . STEINWARW , Z Internationale Atomreaktorbau GmbH, Bergisch Gladbach W. KLEIN Kraftwerk Unio, nAG Erlangen Federal Republic of Germany

Abstract

e demana technicallBase th r n fo do d y clear, safd economian e c reactor systemU GrouKW e p th , consistently used the extensive know-how obtained from German light water reactor technologd an y successful operation of the AVR for development of the HTR-Module. With its operating and safety features and its flexible application possibilities, this reactor concept fulfills current market demand substantiaa o t s l extent.

1. Introduction

After Chernobyl the discussions concerning inherently saf r forgivino e g reactors have been increasingly intensified, not only among reactor safety expert t alsn bu si o the general public. The HTR-Module, developed as a flexible small heat source by Kraft- werk Unio G s affiliate(KWUnA it d )an d company Interatom GmbH, exhibit higa s h degre f passivo e e safet y consequentlb y y usin e inherenth g t safety feature f smalo s l high-temperature reactors.

Detailed information on this aspect will e nexbth et f presenteo lecture e on n i sd . LohnertbDr y . This report deals with the status of the technical design as well

179 e prospecta th se short-ter th f o s d an m long-term market applications.

2. Design Features

The concept of a modular high-temperature reactor with low unit power is a new approac r designinfo h nucleaa g r energy e direcsourcth d indirecr an tfo e t application of process heat in industrial countries (Fig . Furthermor.1) modulae th e r power plant system is well suited for small and medium grid capacities, which are usually require r electricitfo d y generation in developing and threshold countries. However e utilizatioth , f nuclearo n energy in these countries is not restricted to electricity generation e futurealoneth n I .,

Fig 1 .Application HTR-Module th f o s e

180 increasingly vital facilitie e drinkth r -fo s ing water supply or for irrigation could be realize y meanb d f desalinatioo s n plants combined with HTR-Module plants .A furthe r potential application is the extraction l frooi m f o oilsan d residuean dl oi n i s deposits by flooding these with process steam (so-called tertiary oil extraction).

In this case the coupling with hydrogen production coul e interestingb d e th n I . longer term productioe th , f hydrogeo n n and synthetic fuels from coa r naturao l l gas using nuclear process heat will gain importance for countries with domestic fossile resources.

Maximum flexibility regarding plant power ratin d typ an f gapplicatioo e s achievei n d by combining standardized modular reactor units; therefore this system permits step- by-step adjustmen e increasinth o t g energy deman r Thirfo d d World countrie y simplb s y adding on more modules to existing plants. These units operat paralleln i e , giving high overall plant availability.

e maiTh n design criteri d desigan a n elements of the HTR-Module are (Fig. 2) :

Extensive referenc e constructioth o t e n and operational experience wite th h 40 MJ/ R experimentasAV l power plant.

Extensive use of KWU light water reactor technology n particulaI : r this includes the use of steel pressure vessels - under quite comparable operating conditione th wels a s a l- s

181 •j 09400 mm 1 Pebble Bed 2 Pressure Vessel 3 Fuel Discharge 4 Small Absorber Balls Reflecto5 d rRo 6 Fuel Loading 7 Steam Generator: Pipe Assembly Oute8 r Shroud Fee9 e n i dL 10 Live Steam Line 11 Blower 12 Hot Gas Duct 13 Surface Cooler 14 Insulation

2 CrosFig. s Sectio Modulaa f no r Unit with Steam Generator (Primary Circuit)

182 adaptio f techniqueno processed an s s for in-service inspections, maintenance contro d instrumentationan l . Last but not least, the buildings, their arrangement and the secondary and auxiliary systemn accordanci e ar s e with the technology proven in the cours f ligho e t water reactor construction.

Standardized well-proven graphite balls with low enriched uranium as fuel elements (pebble bed core). The waste disposal managemen bases i t d on direct final storage without reprocessing.

By selectin a relativelg y small core diameter of 3 m, the control rods and absorber e locatee b th n n ca si d graphite reflector. This permits gravity insertion of the rods and absorber balls.

Due to the small core diameter in combination with a low core power density possibls i t i , o dissipatt e e the decay heat from the reactor core in a passive way to 3-fold redundant surface coolers located outside of e reactoth r pressure vesse heaa vi lt conduction and radiation in the event of failurmaie th n f o ehea t sink. maximue Th m fuel temperature does t exceeno d 1.60 C durin0° accidenl al g t conditions. Thi n turi s n precludes a significant failure of fuel element particles and a high release of fission products froe fuelth m .

183 Active, redundant primary cooling loops, with associated redundant secondary cooling chains and emergency power supplie n thereforca s e omittedb e .

The aforementioned boundary conditions lead to a power rating of 200 MJ/s per module for the steam generating plant with 700 °C core outlet gas temperature powea o t r d ratinan , g of 170 MJ/s for the process heat applicatio C corn° wite0 outle95 h s ga t temperature.

Reactor core and steam generator are installed in ferritic steel pressure vessels of the proven LWR-quality in side-by-side arrangement with lowered steam generator. Due to this arrangement thermohydraulic decoupling of the reactor core and the operational heat sink is realized after disruption of activ s circulationga e .

Thermal e stealoadth mf o s generator due to natural convection are therefore inherently avoided. Furthermore direct water ingress into the core in the case of a tube failure is not possible.

The arrangement of the components results in easy accessibility for inspection, maintenanc d repaian e r and to low dose rates.

e Chernobye lighth Ith n f o t l catas- trophe, some of these features, such as limitation of the fuel temperature and passive decay heat removal, appear

184 e especiallb o t s tu o y importanr fo t nuclear power plants intended for sites in highly populated or industrial areas. In the extreme case, the HTR- Module plants could even be left without interference and would not have any detrimental e environeffecth n o t - ment. As a result, administrative measures (evacuation etc.) after accidents are not required.

3. Technical Layout of the Nuclear Steam Generation System

e arrangemenTh modulae th f to r reactor unit as a nuclear steam generation system is shown in Fig. 2. Cold helium with an inlet temperatur C enter° e 0 th s25 f o e core from the top via reflector boreholes. Flowing downwards throug e core th hth e helium is heated up to 700 °C. It passes the bottom reflector which is designed as a mixing chamber and flows from the hot gas plenum into the horizontal hot gas duct leading to the steam generator. To achieve upflow evaporation, the hot helium flows from the top of the steam generator to the bottom. From there the cooled-down helium flows upwards through e annulath r cols sectioga d n aroune th d steam generator e blowebundlth o t re placed abov e steath e m generator.

The blower pumps the coolant back to the e reactoannulue horizontath th a n i rvi s l pressure vesse d througan l h channeln i s the reflector to the cold gas plenum on coree th to.f o p

185 Feed-wate d steaan r m flow upwards through the heat-exchanger tubes of the steam generato n counterfloi r e heliumth o e t wTh . live steam produced in this way has a temper- o pressura t d p u an C f ° o e 0 atur53 f o e 220 bar.

The reactor pressure e steavesseth md an l generator pressure vessel together with the connecting vessel forpressure th m e enclosur modulae th f o er reactor. This only comes into contact with cold gas, e selecteth du o s t ductinge ga d , thus per- mitting applicatio e pressurth f o n e vessel technology prove n light-watei n r reactors.

Six absorber rods, which are positioned e boreholeith n e sidth e f so reflector , are provided for control and hot shutdown reactoe oth f r core.

possibll al r Fo e incident e reactoth s r safety system always initiates the same procedure s followsa x sreflecto si e :Th r rods are dropped, the blower is stopped e bloweanth d r fla s closedi e casp th e n I . oa losf f pressuro s e acciden steaa d man t generator heating tube rupture, the steam generator is quickly drained in addition.

In the cold condition, the core is shut f absorbeo y wa dow ry nb spheres, which e containear vesseln i d s abov e thermath e l shield during normal operation. If required, the small absorber spheres are released and drop int 8 elongate1 o d holes provided r thifo s purpose sidth en i ereflector .

186 Each HTR-Module unit is installed in a primary cell e concretth , e wall whicf o s h suppor e reactoth t d steaan r m generator pressure vessels and their internals. Surface cooler positionee sar d aroun e reactoth d r pressure vessel for the removal of heat losses during normal operatior fo d an n decay heat removal when the reactor is shut down.

However, their main task is to protect the reactor pressure vessel and the concrete e reactowallth f o sr cell o environmenta.N l impacts are to be anticipated even if they should fail.

The steam produced can either be used in back-pressura e condensina turbin n i r eo g extraction e generatioturbinth r fo e f no electric power.

The main data of the nuclear heat generating HTR-Module systeth f o m e with steam generator are listed in Fig. 3.

Thermal power MW 200 Core diameter m 3.0 Core height m 9.6 Mean power density MW3 /m 3.0 Helium temperatures (inlet/outlet) °C 250/700 System pressure bar 60 Number of control rods - 6 Number of absorber ball systems - 18 Loading scheme - Reshuffling ~ 15 times Fuel cycle — U.Pu Heavy metal loading of fuel element g 7 Enrichment % 7.8 Burnup MWd/t 80000 Fuel incore time days -1000

Fig.3 Main data for the KWU modular HTR-core for co-generation of steam and electricity

187 The single HTR-Modules of a multi-modular plant can be operated independently. Each module can be individually shut down, run up and connected to the overall plant once it has achieved the operating conditions. Thus e ,cas th f failureveo en a i n f o e single module, operatio e overalth f o nl e continueb plan n a reduceca t t a d d power level.

Fig. 4 shows a cross section of the reactor building, whic s designei h ventea s a d d containment and provides full protection against all external events.

1 Reactor pressure vessel Primar4 y curcuit blower Stea2 m generator 5 Primary cell pressure vessel G Protective shell 3 Connecting pressures vessel 7 Surface cooler

Fig.4 Vertical cross section of the reactor building

188 Fig. 5 shows a site plan of an industrial power plant with two HTR-Modules as an example e reactoTh . r buildin s arrangei g d in replicated serial modular units. The turbine building is located in front of the reactor buildino d housetw an ge th s replicated plant sections of the water system cycle with turbo-se e conth -d an t nected process steam cycle.

Reacto1 r building 2 Reactor auxiliary building 3 Fuel element intermediate store 4 Gas supply unit 5 Wet cells cooler G Switchgear and emergency supply building 7 Turbiné building 8 Hybride cooling tower Offic9 stafd ean f amenitie building 10 Gate building Demineralize1 1 d water tank

I I Nuclear island (NSS Snuclea+ r BOP) I I Conventional BOP

Fig 5 .HTR-2-Modul e Power Statio r Powerfo n - Heat-Coupling, Site Plan e inherenTh t safety feature HTRe th -f o s Module allo e entirth w e conventional steam power plant with all secondary components and system designee b o t smanaged an d d in line with generally applicable guidelines for conventional plant construction.

189 HTR-Module Th Direce 4.th r tfo e utilizatio n of Process Heat

e HTR-ModulTh s alsi e o particularly qualified for the direct utilization of process heat because of its capability to provide heat ahiga t h temperature level.

Dependin e fiel th f applicationo dn go , the steam generato HTR-Module th f ro e generating electric power or process steam is replace helium/heliua y b d m intermediate heat exchanger or by a steam reformer (see steae th ms A reforme . Fig1) . r uses process hea e uppet th onl rn i ytemperatur e range from approx. 950 °C to 700 °C, a steam generator is installed in series to utilize the lower temperature rangC e° fro0 70 m to 300 °C for the production of process steam.

By reducin powee th g r core ratinth e f o g MJ/0 froMJ/0 17 passive 20 mso th st - in e herent safety features of the modular reactor e fullar y retained.

A plant with steam reformer for steam reformin methanf o g describes i e n exampla s a d e of direct process heat utilization (Fig. 6).

5. Barge-mounted energy station

Especially for use in developing and threshold countries, a barge-mounted power plant has been designed as a self-sufficient productioe statioth r fo n electrif no c power or, alternatively, for the co-generation of electricity and process steam, f electricito r o d drinkinan y g water ro

190 1 Reactor Core 3 Steam Reformer 5 Blower Wal2 Stea4 l Coolem Generator r

Fig. 6 HTR-Module with Steam reformer

wate r irrigatiofo r n purposes. This i s especially suitabl r countriefo e s with a low grid capacity and arid areas.

Fig. 7 shows the general arrangement of a 2-Module Plant for electricity and drinking water productio examplen a s a n .

The prelicensed power plant will be completely mounted on a barge in the shipyard for shippin plane th to t g sit y seab e . At the site the barge will be secured in a coastal position or in a prepared lagoon neae coastth r principlen I . barge th , e n als ca e anchore b oe off-shor th n i d e area afloatina s g energy station.

191 Reactor Building

Machine Building Switchgear Building RO-Drmking Water Production

Auxiliary Building 2nd Ste pBrackis- h Water 1st Ste pSeawate- r Gas Control Hoom D Purifi Computer Room cation Electronic Cabinet Plant Conduit Room Gate-] Control wa i yPane l arbonlzation Filter He and / I Startina Diesel ,' i N Waste stor ' Systema s Mam Cablo Ducts Pumps I

Reactor Steam Generator

Tt] o o o o o

Fig 7 .Genera l Arrangemen a 2-Modul f o t e Plant for Electricity and Drinking Water Production The convincing features of this concept are simplified project management, workshop manufacture, cost and time advantages due to préfabrication, standardization and hence possible unificatio e licensinth f o n g procedure.

Significant economic advantages result from the combination of electricity productio d desalinatioan n n usin e reversth g e osmosis (RO) technique: Depending on the conditions, different product mixing ratios cae realizeb n a flexibl n i d e manner. Fig8 . shows a simplified flow scheme of such a plant. Hereby part of the electrical energy generated by the turbo-sets is e desalinatioth brancher fo f of dn plant. e maiTh n consumer high-pressure th e ar s e pump overcominr fo s osmotie th g c pressure and coverin e pressurth g e losses.

HTR-Module200MWth HTR-Module

SteaW M m2 Generato12 r Steam Generator Internal Consumption W 12M

Seawater Seawater

Reverse Osmosis i—— Plant

V 100.000 mVd Drinking Water Fig. 8 Simplified Flow Scheme of a 2-Module Co-Generatioe th Plan r fo t f Electrio n - citd Drinkinan y g Water

193 A further interesting field of application is the generation of high-pressure steam for tertiary oil extraction or the combination of ^-production from oil-gas (wit steaa h m reformer d steaan ) m production for injection in the oil field (with a steam generator connected in series).

6. Economics

In order to underline the expectations for a potential, economic application, e examplr analyseon ou f o r eelectricit fo s y generation alone is offered by Fig. 9 as a suitable comparison measure. The calculatio e economi bases th i n n o d c conditions and parameters prevailin Germanyn i g .

It shows the results of an economic compariso nucleaa f o n r power plant with 4 respectively 8 HTR-Modules and a hard- coal-fired power plan f equao t l size.

DPf/kWh , 16 -

Inflation rate: 2.5%/a Mean interest rate: 6.4%/a 14 - Taxes: 2.6%/a Fuel cycle costs: 0.011 DM/kWth Rate of increase: 3.05%/a Cost of coal: 210-260 DM/t coal equivalent Rat f increaseeo : 2.5%/a 12 — Rate of utilization: 7000 h/a 210 DM/t coal equ. Year of commissioning: 1993

HTR-Module 10 -

I I 800 1600 MWth

Fig. 9 Power generating costs for hard coal power stations v.s. modular HTR's averaged over a 20-year rateoff period (in 1986 currency) (as per 8/86)

194 hard-coal-firee Th d power plant cale ar s - culated for different coal prices

DM 260.—/t (German coal price)

DM 210.—/t (mixed price of imported coal and German coal)

The other main input data for the economic analyse e indicate sar e righ th tn o dsid e of this figure.

From this concludee b analysi n ca t di s that, in Germany., power plants with an HTR-Module can easily compete with fossil- fired power plant equaf so l size. This evaluation improves significantle th n i y application field of co-generation of heat and electricity r othe.Fo r countries- ,es pecially countries of the Third World, this result has to be proved case by case.

Fig. 10 shows the costs for the production of drinking water using a barge-mounted energy station wit selectea h d electricity/drinking water ratio accordino t g Fig. 8. When the results are considered icurrene n th vie f wo t market situation, it is evident that the application chances of such e evaluateplantb n ca s s beinda g ver potentiae y th goo n i d l user countries.

With referenc financinge th o et , plants with a low power rating and hence lower investment costs are particularly interesting, as they can be financed in a simpler way. In the case of the barge-mounted plant, the fundamental possibility of remobili- zation reduces the financial risk significantly.

195 Drinking water costs (DM/m3)

Mean interest rate 5%/a Inflationa rat%/ e1 2-Module Mean interest rat 5%/e6 a Inflation rate 2.5%/a 300 - Mean interest rate 8%/a Inflation rate 4%/a

275 -

2.50 - 4-Module

225 -

200 -

1.75 -

1.50 -

Fig. 10 Drinking water costs averaged over a 20-year rateoff perio n 198(i d 6 currency, without 10/86r taxespe s ))(a

Aspects of realization

From KWU's poin viewf to e technologth , y of the HTR-Module plants has reached such a status that application for the generation of electrical powe d procesan r s steap (u m possibls i ) °C oncet ea 0 e applit53 o.Th - e catiochemicath n i n l industr r similao y r energy-intensive industries offere th s best conditions in industrial countries. Today, an immediate application is also possibl d economian e r tertiarfo c l oi y extraction, especially in combination with the productio f hydrogeno n r developinFo . g and threshold countries 'Seawater desalination and electricity generatio mose th t e ar n promising applications.

196 e detaileTh d plannin d licensinan g g procedure wil e carrieb l yea3 t durin- r ou d2 a g phase according to the prevailing local conditions. In this context it is noteworthy that an advisory panel of the Federal Minister of the Interior has already evaluated and fully confirmed the safety concept. Furthermore, the panel has emphasized its particular suitability for industrial site d sitean s s near townse .Th next ste f realizatioo p s beeha nn initiated wit e applicatioth h r licensinfo ne th f o g technical and safety concept in 1986. The construction perioHTR-Moduln a f o d e power plant (land-based or barge-mounted, in this case including towing to site) is month8 4 3- 6 s dependin n sizeo g . This corresponds to the construction period ocoal-firea f d power plan f equivaleno t t capacity.

Especially after Chernobyl the introduction of nuclear power plants intmarkee th o t presupposes acceptance by politicians and the general public. HTR-Module plants with their high degree of inherent safety features see o havt m vera e y good chanc r sucfo eh public acceptance in the future.

197 SMALL NUCLEAR POWER PLANTS: 10 MW GHR GAS-COOLED HEATING REACTOR

H. SCHMITT Hochtemperatur-Reaktorbau GmbH, Mannheim, Federal Republic of Germany

Abstract

simplea s i ThR , eGH saf economid ean c heating reactor that is designed on the basis of the proven technolog Gas-Coolee th f yo d High-Temperature Reactor. maie Th n benefits derive directly frofavourable th m e technical and safety related HTR-characteristics. These are

low radiation exposure - high inherent safety /evene th eve tn ni of hypothetical accidents few active systems

HRB/BBC-desige Th makeR f GH o wid sa e e th eus f no develope proved an dcomponentsR nHT , sucs ha

proven fuel elements containing coated particles for an extremely favourable fission product retention heliu coolana s a md cerami tan c structures that insure stabilit structuree th f yo s under al1 conditions components manufactured for AVR and THTR and testeservico t p u d e life loads.

The reactor pressure vessel intermediate ,th e circuit, the control room and all safety related system e locatesar d underground which provides protection against all external impacts, sabotage etc. e remaininTh g overground buildin f smallo s i g, conventional desig onld nan y houses auxiliary systems.

199 The whole nuclear heat generating system (primary system) is integrated into a prestressed concrete reactor pressure vessel composes i t I . f o d the core absorbee ,th r rods circulato,a heaa d tran exchanger to transfer the heat via the intermediate circui consumee th o t r circuit.

reactoe Th r control philosophy decae ,th y heat revomal systems and the fission product retention barriers follow HTR-principles, whicwhola s a he characterizes the GHR heating plant by the following features

- simply operating passive safety - less maintenanc operatind ean g effort verradiatiow lo y n exposure - fuel elements for long operating periods low energy generating costs shop préfabrication short construction period.

Plant Design

From 1982 on first analyses of the heat market were conducte Switzerlann i d d abou applicatioe tth n e nucleaoth f r energy from small reactorso t e .Du the specific Swiss residential structure a nuclear short-distance heat supply rated0 5 betweed an 0 n1 MW power output was envisaged. The relevant annual and daily heat requirements are presented an Fig. 1.

This initiative received a positive echo from the public as well as from industry. The preparatory project on the basis of BWR technology was called SHR. Two further projects joined: a second LWR type operating without active components under hydrostatic pressur cooles e ga (GEYSERe d th heatin d )an g reactor

200 THERMAL POWER % 100 20 MW

10 MW

1000 3000 5000 """" 7000 ——— 8760 Operational hours per year Annual heat requirement

Thermal power 100

80 70 / 60 y \ 50 Winter 40

20 Summer 10

0 0 2 4 6 8 10 12 14 16 18 20 22 24 Hours Daily heat requirement

. AnnuaFIG1 . daild an ly heat requirements.

concept (GHR) describe thin i d s paper maie .Th n e listear desig Fign R i dGH . .2 ne th dat f o a

This concept benefits froe safetth md an y operating HTR characteristics which also results in an economic solution. Thus the GHR heat generating costs are competitive with those of conventional heat supply systems.

201 Primary System (Helium Circuit): Thermal power 10 MW 3 m 5 Core volume Power density 2 MW/m3 Numbe sphericaf ro l elements 28.350 Inlet/outlet temperatures 250/45C ° Q 0 1 kg/s Mass flow Pressure 15 bar Circulator power 200 kW

Secondary System (Intermediate Circuit):

Temperatures 95-13C ° 5 Pressure 10 bar 0 6 kg/sWater flow rate

District Heating System (Consumer Circuits):

Temperatures 60-12C ° 0 Pressure 15 bar Water flow rate 40 kg/s

Mai— nR desigFIGGH . .2 n data.

The nuclear heat generating system (primary system integrates i ) d int prestresseoa d concrete reactor vessel equipped with a gas-tight liner, see Fig. 3. The liner is anchored in the concrete. The primary components are the reactor core and internals, the primary coolant circulator and the liner cooling system primare .Th y circui operates i t d with helium cold s an ga d s ga pressura t t ho bar5 a 1 e f .Th eo respectively, °C 0 25 temperatured an .0 45 e sar e hea Th transferres i t d froe primarth m y system o intermediattw e toth e circuit linee th ra scoolvi - ing system i.e primare th . y heat exchanger e line.Th r cooling system is operated at 95 °C inlet and 135 °C outlet waterside temperatures wit systea h m pressure of about 10 bar. The heat from the intermdiate circuit transferres i sr severa o e on l o t consumed r circuits (local heating systems) operating at a water supply temperature of 120 °C (at 100 % load) and returC 6° 0 n temperaturr pressura ba t 6 a e1 f o e (cf. Fig. ) 4 .

202 Circulator

Absorber rods

Ceramic structure PCRV

Thermal • shield

Liner with cooling pipes

reactoR GH rW pressurM 0 1 FIG . e3 . vessel. flexibls i R GH ee witTh h regar theso t d e para- easiln meterca e adapted b y an s modifieo t d d consumer requirements.

The helium coolant flows downward through the core (Fig . Afte3) . r flowing througt ho core e th hth e gadirectes i s d radially outward alon e inneth g r surface of the reactor pressure vessel and then passes upward withi annulae th n r clearance betwee e lineth n r

203 Reactor pressure vessel Outside reactor Outside plant pressure vessel

120 °C 120°C

District heating circuit

Additional heater

15 bar 40kg/s 60°C

gas-cooleW M 0 1 FIG . d4 . heating reactor heat flow diagram.

and the thermal shield before it enters the circulator as a cold gas. An upper gas flow guidance separate e suctioth s d compressionan n e circusideth f -so lator. The cooling pipes are welded to the outer surface of the liner and are grouted in the concrete. The liner cooling system is designed to absorb and conve totae th y l thermal power.

f 28,35o d e cor 0be s Th forme i ea spherica y b d l elements of 60 mm diameter each. The pebble bed is stationary, i.e. elemente circulatet ,th no e ar s d during reactor operation e fue.Th l loadin- de s i g signed for the longest possible residence time, i.e. 14 years.

e temperaturTh e coefficien strongls i t y negative, which results in a stable neutron kinetic plant behaviour.

e reactoTh enrichew operates i rlo a n di d fuel cycle (LEU) e fue.Th l useprovee th s n TRISO-coated UC^-particles.

204 The core is completely enclosed by a graphite reflector wit minimua h side Th me . thicknesm 0 1. f so reflector is equipped with vertical boreholes to accommodate the absorber rods.

The radial circulator is of vertical design equipped with active magnetic bearings. It is electrically driven. The circulator unit is integrated into the top plate of the PCRV.

The prestressed concrete reactor vessel, the intermediate cooling loops including the components provided for decay heat removal as well as the safety-relevant electrical equipmen d instrumenan t - tation including the control room are located below ground. They are protected by a plate which is safe against effects froexternal al m l events.

The entire part of the building arranged above grounf conventionao s i d l design.

The reactor building is equipped with a ventilation system. Using this system various levelf o s negative pressur maintainee ear e differenth n i d t plante roommonitoreth s i wast e f Th .so r eai r fo d acitivity before it is released to the environment throug ventilatioe th h n stack.

Shutdown Concept provides i R GH de witTh o shutdowtw h n systems consisting in total of 24 absorber rods accommodated in boreholes in the side reflector.

The first shutdown system consist rods6 f so . Thehele y theiar yb d r uppe e automatir ar endd san - cally release reacton o d r scram. They drop inte oth reflector by gravity at a rate of about 10 cm/s. This system is capable on its own to take the reactor to

205 a safe condition under all events of maloperation and design basis accident sprolongea alsr fo o d period.

The second shutdown system comprises the remain- incontro8 1 g l rods. Thee subdivideyar d into three groups. 6 of these rods each are additionally provided for trimming and control.

The rods of the second shutdown system are also automatically released in the event of reactor scram. In that case they drop by gravity at a rate of about cm/s1 .

e secon th e rod Th f do s shutdown systee ar m equipped with a drive system which is diverse from e firsth tha tf o t shutdown system.

Withdrawal of the rods of the second shutdown system is only possible if all rods of the first shutdown system are in their upper end positions.

The second shutdown system is more effective thae firsth n t shutdown system e secon.Th d shutdown system is capable to take the reactor to a safe condition under all events of maloperation and design basis accident prolongea r fo s d period.

Decay Heat Removal Concept (Fig) 5 .

decae Th e storey th hea d dremovee t an b hea n tca d e followinth f o e gbon y systems:

- Plant main cooling system using the local heating system as a heat sink:

When operabl maie th en cooling syste uses i m d r operationafo l decay heat decar removafo yd an l heat removal in the event of accidents. No specific requirement howevee sar availabilitys r it mad o t e .

206 Reactor pressure vesse | lOutsid e reactor pressure vessel

Hit j—A-

-1——cx> Decay heat removal system

Primary circuit Secondary circuit

cooleW M d0 FIG1 heatin . 5 . g reactor decay heat removal system. decae Th y hea removes i t naturay b d l convectoia nvi the intermediate circuit districe th o t s t heating system.

Two-loop decay heat removal system:

In the event of failure of the plant main cooling system the decay heat is removed via two decay heat removal loops with decay heat removal coolers. Each decay heat removal loop is connected to an intermediate circuit.

The decay heat removal cooler consists of a storage tank filled with démineraiized watert .I arrangeo is o ensurt s a de sufficient natural convection for decay heat removal.

Each loop of the decay heat removal circuits is so designed that in the event of damage in the other loop and failure of the external electricity supply the total decay heat can be removed. This appliepressurizee th o t s d reactoo t s wel ra s a l depressurizatio reactore th f no .

207 Cooling of the storage tank is generally effecte y specificallb d y provide r loopai d s when external electricity supploperationn i s i y .

e evenr circulatoIth nai f failurto e th f o er e tanth k coolin effectes i g reducea t a d d pacy eb natural convection of air. In this case evaporation of the water in the decay heat removal storage tank may e lonoccuth gn i rrun e capacitTh .e on f o y storag edesigneo s tan s i k d thae decath t y hean tca be removed for a period of more than 100 hours after e occurancth accidentn a f o e .

Decay heat removal throug e PCRth hV wall:

e extremelth Eve n i n y improbable evenf to total failure of both intermediate cooling loops decae th ye remove b hean ca td throug e walth f ho l the pre'stressed concrete reactor vessel with the reactor pressurized as well as under depressurized conditions.

e relevans temperatureTh ga t tho d heliusan m pressure e presentesar Fign i d . .6

Activity Containment

The concept of activity containment is based on the multibarrier principle

- coated fuel particle - graphite matrix of the fuel element - prestressed concrete reactor vessel reactor building (vented containment).

The first and the second barriers are provided e sphericabth y l graphite fuel elements containing the fuel in the form of coated TRISO particles which are embedde graphita n i d e matrix.

208 500

400- ) Deca(2 y heat remova naturay b l l convection usin auxiliare gth y systems S 300- (3) Decay heat removal by convection and radiatio evene th totaf n to i l failurf eo l auxiliaral y systems 200 0 1 Timh e 0 1, 1 0, 100

temperaturs ga t Ho functioa R s ea GH f timnW o M e0 1

12- (2) Decay heat removal by natural convection usin auxiliare gth y systems 10- ) Deca(3 y heat remova convectioy b l n arRT radiatio evene th f totao n ti l failurf eo l auxiliaral y systems 8 0,1 1,0 10 Time h 100

10 MW GHR Primary coolant pressure as a function of time

FIG. 6 .

This ensures an extremely low release of fission products under all operating conditions. This applies also to accident conditions since the relatively low temperature e fueth l f selemento t resulno a n o i std considerable increas fission i e n product release.

The third barrier is provided by the leaktight failsafe prestressed concrete reactor vessel.

209 For pressure limitatio preventiod an n overf no - stress the prestressed concrete reactor vessel is provided with redundant depressurization systems (safety valves and rupture discs).

The reactor building together with the auxiliary facilities for retention and filtering of potential leakage f reactoo s r coolant represen e fourttth h barrieactivite th f ro y containment.

Due to the low coolant gas activity coolant leakages resulting from accident e releaseb n ca s d environmene th o t t throug ventilatioe th h n channels.

Accident Behaviour

The GHR safety concept makes extensive use of generae th passivR HT l e safety characteristicse .Th high level of passive safety in the GHR is mainly obtaine e limitatioth y b d unif o n t powe powed an r r density to low values so that even in the event of extremely improbable accidents an impermissible releas f fissioo e n products froe corth me must no t e expectedb . 7 e Fig. ,se

FueR l GH Elemen e Th t Design Values and Design Limits

Design values Design limits

Average power 0,6 kW

Maximum power 1,7 kW 4,5 kW

5 7 GWd/t5 GWd/t11 o o Bum-up

Central temperature 684 °C 1.250 °C (<800°C)

FIG. 7. The GHR fuel element: design values and design limits.

210 e eventth Als n si o exceeding design base accidents the GHR offers high safety:

inherene th o t te temperaturDu - e limitation under all accident conditions important time reserves are available to take appropriate measures.

Even in the event of severe accidents access reactoe tth o r buildin possibles i g .

Even in the event of total failure of the decay heat removal system the core temperatures remain relativel w limitinlo y e releasth g f o e fission products.

Even in the event of total failure of the decay heat removal systems it is possible to remove the decay heat throug vessee th h l walls. Under such conditions the maximum core temperature of about 850 °C (locally depressurizea n i ) d reactor remains low. w temperaturelo e e corth th e o d t structur an se Du e composed of ceramic materials core damage and core meltdow e absolutelb n ca n y ruled out. Thers i e onl transiena y t releas f iodino e whic1 13 e h results

1. Normal operation 0,2 mrem/a Normal annual release

5 mre0, m 2. Refueling

3. Depressurization accident

Without failure of 3,0 mrem forced circulation

With failure of 6,5 mrem forced circulation

FIG. 8. GHR — Radiation exposure to the environment (effective whole bod distance)m y0 dose10 t .(a )

211 imaximua n m emissio maximue f 0.0Th no . 3mCi equivalent whole body dose resulting from the release of radioactive iodin d radioactivan e e noble gases amounts to not more than 6,5 mrem (immersion, ground radiation and inhalation), Fig. 8.

This favourable situation 6,5 results in a high flexibility with respec o sitt e selectionR .GH plants can be sited near urban centers.

212 HT INDUSTRIA0 R10 L NUCLEAR POWER PLANT FOR GENERATIO HEA F ELECTRICITND O TAN Y

. BRANDES S Hochtemperatur-Reaktorbau GmbH W. KOHL BrownG A ,e BoverCi d un i Mannheim, Federal Republic of Germany

Abstract

Nuclear power plants operated with small high- temperature reactors provid economicn ea , environmentally favourabl d reliablean e electrit head an yt s-uppl r industriafo y l plant d communitiessan .

Based on their proven high-temperature reac- witr to h pebbl cored ebe , BBC/HRB have developed an HTR 100 plant which combines favourable capital costs and high availability. Due to the high HTR- specific standards this plan expecialls i t y well suite sitinr fo d userd g en nea.e rth

The safety concept is designed as to permit further operatio plane th deca r to f n o y hea- tre moval via the operational heat sinks in the event of maloperation and design basis accidents having a higher probabilit occurencef yo . Thus high loads avoidede ar , especiall reactoe th n i yr pressure vessel. In the event of hypothetical accidents the decay heat is removed from the reactor pressure vesse radiationy b l , conductio convectiod nan o nt a concrete cooling system operatin naturan i g l convectio. n plan0 10 tR concepHT w ne exampln Ata e a s th f o e twin-block plant design for extraction of industrial steam is presented.

e reactoTh r buildin s locatee centei g th n i rd overale oth f lsafeta plants i yt I .containmen t

213 building with its inner structure detached from the outer shel r saffo le controy externaan f - o l im l pacts. Its inner structure is designed identically foo reactortw r units nucleae .Th r steam supply system syste s includei m steea n i dl reactor pressure vessel in which all the primary components e integratedar e safety-relevanth l .Al t systems are installed in the reactor building.

e buildinTh g configuratio e planth s f i tno characterized by clear functional structuring, short distance systemr fo s personneld san , economic design, short construction period and low space requirement.

1. Introduction

e earlIth n y 8Oies smal d medium-sizan l e reactors gained more attention. Especialle th r fo y suppl f heao yd electricit an t o industriat y l fac- tories only small power plants are suitable. Since 1980 subsidiare Browth d n an Bover e y Ci compan& i y Hochtemperatur-Reaktorbau GmbH have developed a small High-Temperature Reactor designed for 258 MWth thermal power and a net electrical output of 100 MW concepR AV (HTR-100 e makind basie th tan th f so n g)o use of the advanced technology of THTR for component systemsd an s .

This reactor type is particularly suited to supply electric power and heat for application in industry, communities and industrial threshold countries. Siting near industrial centers and near centers with high population density are permissible because of the high safety standard of the reactor with passive safety characteristics. To hold cost down special emphasi s beesha n placen o d

214 the greatest possible extent of préfabrication and preasserabl plante th f .o y

Earl 198n i y 5 Developmena t Consortius mwa formed to initiate the construction of a reference plant and optimize the technical, safety engineerin economid an g c aspectHTR-10e th - f so 0de signe y BBC/HRBb d .

Members of this consortium are:

Hochtemperatur-Reaktorbau GmbH (HRB) Dortmund Brown ,G (BBC A Bover e Ci ) i& Mannheim Deutsche Babcock Maschinenba G (DEMuA ) Ratingen Mannesmann Anlagenbau AG (MAB) Düsseldorf Strabag Bau-AG Cologne INNOTEC Gesellschaft für Spitzen- technologien und Technologietrans- Essen . EnergietechniCo & feH rmb G K k

with INNOTE centrae actinG CK th s a gl coordinator.

. 2 Nuclear Steam Supply System

The nuclear steam supply system of the HTR-100 consists of the reactor core, the steam generator, the circulators and the shutdown facilities, all integrated withi steea n l reactor pressure vessel (FIG. 1). The reactor vessel is located in the e reactocenteth f o rr building surroundea y b d cylindrical biological shield. The inner surface biologicae oth f l shiel equippes i d d wit concreta h e cooling system to protect the concrete from high temperatures (FIG. 2).

In normal operating condition reactoe sth r core is cooled by an upward coolant flow at a pressure of 70 bar. The hot gas leaves the core at a temperature o, flowf °C abou s0 p througreflectot74 to e th h r and mixes with various bypass flows which results

215 Section A-B

Circulator

Feed water

Steam generator

Container for small absorber spheres Control rod.

Top reflector

Side reflector

Reactor pressure vessel

Thermal side shield

Thermal bottom shield Support structure Fuel element discharge equipment

SectioD nC-

Fuel element discharge pipe Control rods Graphite reflector buttress

: Reacto1 FIG. r pressure vessel with internals

216 Spent fuel Reactor Reactor Connection Turbine Process storage auxiliary building building hall steam building supply J______L + 44.00m

+ 24.0m

+ 9.0m + 0.0m

-11.5m

Reacto1 r pressure vessel 6 Turbine 84300 2 Reactor 7 Generator Stea3 m generato, r Contro8 l room 4 Circulator f* f ' ^. 9 Hot workshop 5 Biological shield "ï. i~?'£ , 10 Fuel cask storage Access building Electricial equipment building -147450-

industria0 10 R HT FIGl: 2 nuclea. r power plant, cross section, longitudinal section

in a mean steam generator inlet temperature of Durin. 70°C 0 g normal operatio thermana l power of 260 MJ/s is transferred and steam is generated bar0 19 .d an Afte C ° r0 transferrin53 t a heae th gt to the steam generator the coolant gas enters the circulators at a cold gas temperature of 250 °C. The circulators forc coolane eth t - througan e th h nular clearance between the inner wall of the pressur ee stea vesseth md an generatol coree th ,d ran respectively, into the lower part of the reactor pressure vessel where it enters the pebble bed through opening bottoe th n mi s reflector.

217 2.1 Reactor Vessel

The reactor vessel is the pressure containment overale th r lfo primary coolant circuit (FIG. .1) The total height of the vessel is about 30 m. The maximu vessee Th msupportes . i lm diamete 1 6. s di r approximately at its centre of gravity. This permits the control of expansion an shifts resulting from high temperatures. The vessel is manufactured from creep-resistant fine-grain stee MnMoN0 2 l forg5 5 i - ings in compliance with the design and inspection requirements proven in pressure vessel construction.

2.2 Reactor Core

The reactor core is a cylindrical bed of spherical fuel elements of about 3.5 m diameter and a mean height of 8 m. It is embedded into a graphite reflector comprising a bottom reflector, side re- flectop reflectorto d ran unie verticall.Th s i t y mounte star-shapea n o d d metal support, whics i h itself positioned on the support ring of the reactor pressure vessel. This permits thermal expansion and prevents torsional effects side .Th e reflector s interlockei d witthermae th h l side shield.

The reactor cor s fuelei e d with spherical fuel elements of the TRISO type with multiple passage of the spheres through the core using the same type of fuel elements with low enriched fuel which have Juelichn i beeR AV e fuen .Th e testel th element n i d s are introduced continuously into the core through five refuelin p reflectorto g e pipeth n .i s They pass downward through the reactor core before being finally withdrawn through four fuel element discharge pipes located suppore belo th core n th wi e t structure. A burnup measuring facility equipped with a computer determine e fueburnue th th sl f elementso p .

218 TABL : ReactoI E r Core Main Data

Diameter 3.45 m Pebbl heighd ebe t 8 m Numbe sphericaf ro l elements 317 500 Power density 4.2 MW/m3 Coltemperaturs ga d e 255 °C Hot gas temperature 740 °C Helium pressure 70 bar Helium mass flow 111 kg /s Initial core U235 initial enrichment 6 % Heavy metal content per fuel element 8 g Graphite elements 55 % Equilibrium cycle U235 initial enrichment 9 % Heavy metal conten r fuetpe l 14.5 g element Graphit elements 45 % Residence time of fuel elements 977 days (full power) Mean burn up 99.4 GWd/t Heavy metal content of the core 2.36 t

Fuel element spheres with a high burnup are discharged from the cycle and are replaced by fresh fuel elements, while fuel elements which are not yet completely spen pneumaticalle tar y recirculated into the reactor core.

reactoe Th r cor TABL( e havin) I E meaa g n core power densitMW/m2 4. f 3 o yconsist 31750f so 0 spherica ldiameter) n i element m m e equilibriu th 0 (6 n sI . m cycle only 55 % of the spherical elements are fuel elements, the remainder are graphite elements.

A ta loameae th dn 9 factoresidenc0. f ro e time e fueth l f o element equilibriue th n si m cycls ei three years. This amounts to five passages through

219 mea a e cord nth ean burnu apprf po 0 GWd/.10 f to heavy metal two-zon.A e cor establishes i e usiny b d g different burnup ages in both zones. The initial enrichment is 9 % U235.

The initial core consists of a mixture of graphite spheres, absorber elements and fuel elements with 6 % initial enrichment. The initial core fuel element e replacear s y equilibriub d m fuel elements during the first years of operation.

e reactoTh r cor controlles i e o shutdowtw y b d n systems. 24 absorber rods inserted in holes of the side reflecto usee contror rar fo d d reactoan l r scram r longter.Fo m shutdown small absorber spheres (app. 9.5 mm) are filled into vertical holes of the four graphite buttresses.

8 absorber rods are used for load control. They compensat necessare th e y excess reactivit alloo yt w load-following between 50 and 100 % load. FIG. 3 shows lateral power distribution fult sa l powen i r

Upp«r Core Part

Boundary Inner/outer Core

Side Reflector Lower Core Part Bore Holer sfo Absorber Rods

Core

45 Bore Hol r efo Buttress MWMl3 Small Absorber Spheres : FIGLatera3 . l power distributioe th n ni reactor core

220 the upper and lower part of the core (equilibrium cycle). The control rods are only inserted in the upper part of the core to compensate the excess reactivity. The figure shows also the boundary be- outetweee th e inne rth nd an rcore .

FIG show4 . radiae th soutles ga l t temperature distribution. The maximum gas outlet temperature exceed . e averagK th s5 2 onlC ° y e b y 0 valu74 f o e

800- TGBS - M- 770- 760- 750- 740 Mean gas outlet temperature 730- 720- 710- 700-

680-

05 1.0 1.5 1.725

RwSutQn]

FIG. 4: Coroutles ega t temperature distribution

2.3 Shutdown System

For trippin reactor.1e th g 6 absorber rodf so THTR design are automatically inserted into bores in the side reflector. The reactivity value of these rods is at least 3.4 % A k for compensating reactivity effect unlimitef so d water ingress. Under normal condition core th se temperature after reactor tri reduces i p d onl y approximatelb - y en o t K 0 10 y sure a quick restart.

The small absorber spheres, providing the second shutdown system (FIG. 5), are manually released from containers into the channels in the buttresses.

221 _ Small absorber spheres-storage vessel

Supply channels Release mechanism

Ring Delivery cone

SAS-RII|ng_

Return pipe

Exit-valve

Feed chamber Cross section A-B

FIG. 5: Small absorber spheres shutdown system

The small absorber sphere shutdown system is used, when shutdown periods are longer than one day and for long-term shutdown at ambient temperature.

The container e installear s d outsid reactoe th e r pressure vessel . The release mechanism is activiated releasing the spheres into the channels in the graphite buttresses by gravity.

For re-startin reactore th g boroe ,th n spheres are recirculated to the containers under gas pressure via exit valves below the core, feed chambers and return piping.

The reactivity value of the small absorber sphere syste greates i m r % tha£±K, 5 1 n which compensate maximue sth m reactivity demand occurrinn i g the initia. ) AK l% cor3 (1 e

222 2.4 Internals

The cavity holding the spherical fuel elements is formed by the ceramic internals, the bottom reflector, the side reflector and the top reflector(FIG. 6)

Refueling Inlet tube tube KLAK

Top reflector

Dowel

Bottom reflector

Section H-J Section K-L

Support/fixing point Small absorber sphere shutdown system

: FIG6 Cor. e structure

223 Mechanical loads from dead weight and spheres pressure as well as seismic loads are compensated e graphitbth y e component d transferresan e th o t d adjacent metal component parts.

The rigidity requirements of the graphite structur vert no y e strictear , since jamminf o g fuel elements canno tpebble occuth d corn i be ere and flowine sphereth f o gensures si d eve casn i n e of irregularities in the reflector structure. The side reflector is formed by an inner and an outer cylindrical wall consistin f reflectoo g r blocks o forpilet mp u dcolumns e block.Th e verticallar s y fixed by dowels, horizontally they are connected by keys keye .Th s simultaneously serv s sealsa e , limiting the bypass flow through the vertical gaps betwee columne permissible th nth o t s e amount.

The four graphite buttresses protruding into the reactor core are connected with the side reflector.

The top reflector consists of layers of canti- levered segment-shaped graphite blocks and a central plug e cantilever.Th e supporte sar e inne th ry b d layer of the side reflector.

For the gas flow and for the addition of fuel elemente smalth ld an sabsorbe r spheres suitable hole providee e segment-shapesar th n i d d blocks.

2.5 Steam Generator

The hot gas leaving the core flows upwards e steaith n m generator transferrine th s hea it o gt feedwater (FIG. 7).

watee Oth n r side steae ,th m generator consists o tubotw fe bundle abouf so heatin0 16 t g tubes each,

224 13700

Feedwater FW

FIG. 7: Steam generator

manufactured from Incoloy 800, which are wound into each othe helicoilsn i r e coolanth n sides .O tga , bundleo thtw e s forunite on m transienn .I t condition wels sa durins a l g startu d shutdowpan n pro- cedurefeedinn i d e tubsan on g e bundle onlye ,th steam generator tubes are never exposed to impermissible load.

e feedwateTh r inlet maid san n steam outlete sar located laterally at the upper cold end of the steam generator pipee .Th s penetrat reactoe eth r présure vessel horizontally immediately abov steae th e m generator.

225 2.6 Circulators

After transferring the heat to the steam gener- atoheliue rth m coolant circulatoro entertw e sth s a cola ttemperaturs ga d C (FIG° abouf e o . 0 8) .25 t

4300

1 Circulator shaft 2 Rotor 3 Motor 4 Radial magnetic bearing 5 Axial magnetic bearing Catche6 r bearing 7 Shutoff valve with actuator

FIG: Heliu8 . m circulator

The coolant gas circulators are single stage radial blowers drive speed-controllea y nb d asynchronous motor. The circulator shaft is held in active magentic bearings so that there is no me- chanic contact between rotatin statid gan c parts. The bearings are free of wear requiring no lubri- cants.

7 Fue2. l Circulating System

The fuel circulating system permits continuous loadin withdrawad an g f spherelo s during reactor operation. The greater part of the fuel circulating system is located below the reactor vessel. The flow diagram (FIG. 9) shows the arrangement. The

226 Pressure Fuel balancing tanks discharge Multiple switch Q

I Loading block ..Pneumatic Yelevation i system Distributor block --o-- Collecting Burn u Damaged block measuring Discharge «-*-! spheres device block container Legend: Pressur0 e barrier ® Switch/collector Sphere0 s release valve O Counting device

FIG. 9: Fuel circulating system

spherical elements - fuel, graphite and absorber element deliveree ar powee s- th ro t dstatio n ni transport casks. The casks are transported from the fresh fuel storage fuelinth o et g station. From here they are transferred to the distributing station and then forwarded pneumaticall pipe5 core a th sevi o t y allowing to build up a two-zone core.

After passing throug core gravityth hy eb e ,th elements leave the core at the bottom and are directed to the singulizer which is coupled with a separato damager fo r d spheres. Damaged spheree sar eliminated. Intac burnua o t pd sphere measle e -sar urement device. Spent fuel elements are discharged. Fuel element st full no whic ye - ar hspentre e ,ar circulated into the core.

2.8 Helium Purification System

The helium purification system removes chemical

impurities such as H, CO, H0, C0 using molecular 2 2 sieves to minimize corrosio2 n of the core graphite.

227 Graphite dust is also removed. However, neither noble gases nor nitrogen are removed. The molecular sieves are regenerated with helium.

3. Safety Concepts

1 Shutdow3. n

Reactor scram is automatically achieved by absorber rods which are inserted into holes in the side reflector by gravity. The reactor remains in subcriticat ho l conditio least . a h 0 r t2 n fo

r longterFo m cold shutdow smale nth l absorber sphere e releasesar d - intre channele oth e th n i s flector buttresse gravityy sb .

3.2 Decay Heat Removal

After reactor scram the decay heat generated in the core has to be removed. For this purpose two independent and diverse systems are provided in the HTR-100.

During shutdow powee th rf no plan t from normal operatio nuclea e powee th nth f ro r steam supply e steam-feedwatesysteth d an m r circui slowls ti - re y duce The. abouo % operatioe t dn% th 0 t1 fro 0 10 m n is switched to one of the two loops of the operational startup and cooldown system (FIG. 10) by openin s shutofit g f valve closind san maie th gn steam shutoff valve. The wet steam is condensed in the steam condenser of the operational startup and cooldown syste auxiliare cooled th an m y b d y cooling water system in section-type coolers. The condensate is recirculate steae th mo t dgenerato r througe th h feedwater line.

228 Steam/feedwater circuit Process steam circuit

127 MW bar0 19 , 530<>C kg/3 20 ,s bar, 2700C

Gross electrical output 175-10W 0M Process steam mass flow 170-50h 0t/

FIG. 10: HTR 100 industrial nuclear power plant heat flow diagram Under upset conditions, and under most accident conditions resultin reacton i g r scram heae ,th s ti removeauxiliare th y b d y feed-water system.

The auxiliary system including the cooling chain, has builtin redundancy. The steam generator is designed with two loops and each reactor is equipped with two circulators. An additional auxiliary power suppl provides i yo diesels tw y b de .Th measures result in a high availability of the auxiliary feedwater system. This system e b alon n eca use removo t d e deca ye even th hea f upsen to i t t condition durind san g most accidents resultin reacn i g - r scrato m without undu componentse e th loa o t d .

Accidents suc turbins a h e scram pipr ,o e rupture are completely controlled by the auxiliary feedwater systemevene th f failurte o n circulator.I on f o e ,

229 plant operation is continued at reduced power until a scheduled shutdown e .even th Eve f failuro n ti n e of both circulators, heat transfer from the reactor e steath mo t generato naturay rb l convection within the primary circuits is sufficient for heat removal usin auxiliare th g y feedwater system.

Depressurization is controlled by one circulator. In case of failure of one steam generator loop, decay heat is removed by the second loop.

The high heat capacity and the low power density HTR-10e th f o 0 permi tlimitea d interruptio decaf no y heat removal offering timr manuafo e l action.

e rarIth ne accidenn evena f to d simultatan - neous failurauxiliare th f o e y feedwater systee th m hea removes i t d throug e concretth h e cooling system (FIG . Thi1) .s syste independens i m auxiliare th f to y feedwater system heae .Th transportes i t d froe th m reactor cor heay eb t conduction, radiation cond ,an - vection throug outee th h reactoe r th wal f o lr pressure vessel to the surface coolers of the concrete cooling system. The cooling water in the concrete cooling system is heated, flows upward by natural convection and discharges its heat to the atmosphere by evaporation. Cold water flows inte surfacoth e coolers froreservoirse th m - re capacite e .Th th f yo servoirs is sufficient for up to 2 days heat removal.

3.3 Activity Release

During normal operation of the reactor the radiation exposure result smala n i sl environmental dose near 0.01 mSv/y.

hypotheticae Ith n l evendepressurizatioa f to n accident with longterm interruption of decay heat removal by the operational startup and cooldown system, the fuel element temperatures in a small

230 core region rise to a maximum of 1680 °C after about 60 hours for a short period. In this case the hea solels ti y transporte heay b d t conductiod nan radiation to the concrete cooling system outside the reactor pressure vesselconcrete th f I . e cooling t availablesysteno s i m e fueth ,l temperatures will nevertheless remain belo thin wI s168. acciden°C 0 t the radiation exposure at the most unfavourable spot results in

whole body 0.16 mSv (0.3 %) thyroid (infant) 3.4 mSv (2.3 %).

Figures in brackets give percentages of the values legally permitte e Federath n i dl Republif o c Germany unde accidenl ral t conditions.

4. Balance of Plant Description

e forementioneTh e th dusee r b NSS fo dn Sca generatio heatf no , electric power stead ,an m r applicatiofo C ° 0 industrn 50 i n o ut p d an y communities. A single or twin-block station is offered. As an example a twin-plant version for cogeneratio procesf no s stea electricitd an m y is described below.

4.1 Nuclear Island Arrangement

e generaTh l arrangemen e buildingth f to r fo s the HTR-100 power plan shows i t FIGn . i n2 .

The buildings are arranged in a compact configuration with the reactor building in the center. On buildinge e reacto th th e sid on e f eo rar s containing control areas such as the reactor auxiliary building and the spent fuel storage building. On the other side there is the conventional plant, the connection building, the turbine hall and the building housing

231 the industrial steam supply plant. The electrical equipment building and the access and security building are arranged perpendicular to the main axis of the power plant buildings. The overall length of d totae an th , lm , m buildine 2 widt 8 8 th s 14 i h s i g reactoe th r buildin. m heigha 4 4 s f ha gto

2 Stea4. m Power Plant

FIG. 10 shows the major components of the main steam turbine generator condensate/feedwated san r twie systemth n r plantsfo .

e steaTh m supply syste designes i m d according principlee th o t cogeneratiof so orden i n produco rt e industrial steam at 270 °C, 16 bar and at a mass flow togetheh t/ W 0 M 50 r5 - 17 wit 0 - o hf17 W M abou 0 t10 gross electric output.

The main steam from the two generators flows tcommoa o n header from wher high-pressurea e (hp) backpressure turbine is supplied at a temperature of 530 °C and a pressure of 190 bar. The exhaust steam passe mp/la o st p extraction condensing turbine. The overall gross electric output of the two turbosets is 127 MW at 400 t/h industrial steam extraction.

The exhaust steam from the mp/lp turbine is condensed in an air-cooled condenser. After passing throug three-stagha e regenerative feed water header, the condensate is divided between two loops and fed o deaerator-heatintw f o inte oon g stages e feed.Th - wate thes i r n pumpe o stead tw intme oth generator s through the hp preheaters.

To achieve high availability e industria,th l steam supply plant is also designed as a two-loop system condensate .Th e flowing back froe industh m -

232 trial steam networ preheates i k thren i d e staged san converted into steam at 270 °C in a steam converting and superheating station.

The heating steam for the superheater and converter is extracted from bleeds in the turbines. The collected condensate cascades e froe stagth on m o et next, transfer s residuait s l finalls heati d ,an y recirculated intsteam/feedwatee oth r circuite .Th industrial steam circuit is separated from the steam feedwater circuit by the heat exchangers so that no activity frosteam/feedwatee th m r circuit will reach e industriath l steam circui leakaga f ti e occurs.

This two-loop design system permits continuous operatio plane th tf n o even during inspectio- in f no dividual parts of the plant. Thus, for example, one shue b tn dowca inspectior reactos fo n it d ran n purposes, whil secone eth d reactor continue operateo st ,

In the event of failure of the hp turbine, the mp/lp turboset is supplied directly with main steam throug reducinha g station casn . I f failur eo f eo mp/le th p turbine e stea,th mp h expande th n i s turbin bleee th do et pressur industriae th r efo l steam supply plant botn .I h case powesa r output obtaineds i f abouW o M 0 t9 , more than hale th f total output of both turbines. This shaft asymmetric turbine connectio higa s h nha efficienc y during both normal operation and under upset conditions. Compare otheo t d r multi-shaft turbine connections, it requires less turbin e spacth n ei e buildind an g less instrumentatio controd nan l equipment.

The heat flow for the HTR-100 power plant is designe achieveo t d :

High availability of industrial steam and electric power

233 A rang capacitf eo r industriafo y l steam of 170-500 t/h;

Gross electrical output of 100 MW-175 MW;

Constanf o t h supplt/ t leasa 0 f t17 yo industrial steam and 80 MW of electrical power;

percen0 9 o Uefficienct t p tne r fo y hea powerd tan .

234 BASIE TH C— DESIG0 50 COMMERCIAR R NFO HT L HTR POWER STATIONS

E. BAUST, J. SCHÖNING Hochtemperatur-Reaktorbau GmbH, Mannheim, Federal Republi f Germanco y

Abstract

With the completion of the THTR 300 and the development of the THTR follow-on plant HTR 500, BBC/HRB have take pebble n th high-temperatur d ebe e reactothreshole th o t r f commercializationo d . e referencth s i THTe 0 Th e entir30 Rth plan r efo t high-temperature reactor line in the power range be- additionn I . MW e experienc0 ,th 60 d an twees i e0 10 n available whic s beehha n acquired witsuccessfue th h l 19 years operation of the 15 MW AVR in Juelich, also constructe BBC/HRBy b d . Anexa s t ste markef po t penetratioe th n i n Federal Republi f Germano c e constructioth y e th f no HTR 500 is intended, a power station for electricity generation and the possibility to extract process heat and district heat. The power size of 550 MW takes accounmodifiee th f to d e nucleademanth n o dr power plant market which for reasons of grid size, financing expensee loweth rd an sincreas elecn i e - tricity demand internationally shows a trend towards smaller power units in the range between 400 - 600 MW. f HTR-specifio e us e th Duco t e safety characteristics and extensive application of THTR 300 technology, the 550 MW power station is economically competitive with large PWRs. For special application n industri s y (cogeneration of steam and electric power) and for special siting conditions (near industrial centers) the small s beeha nR developeHT W M 0 10 d whic n als conca e hob - twin-planW M 0 structe20 a ts a dunde r economic conditions, competitive with conventional power plants of the same size without involving the problem of air pollution. As a further HTR application, small heating reactors of about 10 MW can be designed on the basis pebble oth freactod ebe r concept. Afte Chernobye th r l reactor disaste hige th rh inherenbees ha nR undeHT t e safetrth discussiof o y n with increased emphasis presene th n .I t situation characterize sensitivy db e attitude toward nuclear energy, the HTR with its high safety potential offers favourable conditions regarding its acceptance by the public and the politicians.

235 decisiva s i ThR eeHT innovatio e fielth f o dn n i reactor technology which represents an important contribution to the intermediate and long-term supply of safe, environmental and economic energy to the entire electricity and heat market. particularls i ThR eHT y suite Thirr fo d d World countries turninpeacefue th o t g l usage nucleaf so r energy for the first time.

. 1 PRESENT SITUATION On Novembe 198, 16 r THTe prototyp 0 5th 30 R e nuclear power plan Hamm-Uentron ti p supplied first electriqity to the public grid, on September 23, 1986 full power was reached. With this the first large high-temperature reactor with spherical fuel elements started electric power generation. The nuclear power plant was constructed by Brown Boveri & Cie AG (Mann- heim) in a consortium with the subsidiary company Hochtemperatur-Reaktorbau GmbH (Dortmund) and the fuel element manufacturer Nukem GmbH (Hanau) on the plant site of the Westfalen power plant of VEW Dort- mund. The plant owner is Hochtemperatur-Kernkraftwerk GmbH (HKG), an association of regional and local public utilities. firse Th Germate th stedevelopmen R f npo HT s twa the construction of a 15 MW electric power experimental reactor in Juelich in the Federal State of Northrhine Westphalia. This reactor, called AVR, was taken into operatio usere th sy n b grou p "Arbeitsge- meinschaft Versuchsreaktor" (abbreviated AVR)n i 1967 and has demonstrated an extremely satisfactory operating behaviou year9 1 r s rfo (Fig . Lon1) .g years of operation of the experimental reactor at a maximum coolan temperaturs havtC ga ° e0 verifie95 f eo e th d generatior fo suitabilitR HT nucleaf e no th f o yr process heat. e seconR linth HT s eda e THTe th 0 ste Th 30 Rf po represents the essential basis of commercialization High-Temperature oth f e Reactoe Federath n i rl Repu- blic of Germany (Fig. 2). It is the reference plant entire forth e high-temperature reactor line witl hal follow-on poweplante th rn si rang e d betweean 0 10 n e prototypth s A . eel includeW M 0 60 s large design reserves, higher power levels up to about 600 MW can be achieved with almost identical overall dimensions and structural volumes. In a twin plant design, a powe coverede b r n o 120rangt ca p 0.W eu M With the THTR 300 the nuclear licensing procedure for high-temperature reactor bees sha n introducen i d the Federal Republic of Germany establishing the re-

236 FIG. 1. 15 MW AYR Nuclear Power Plant Jülich.

Nucleae MW 0 r Powe30 FIG . 2 . r Plant Hamm-Uentrop.

237 quired know-how with experts and authorities. In addition, a competent manufacturer and supplier industry has been built up. Finally, also the competence on the user side has been substantially increased. The commissioning results of the THTR 300 have fully confirmed the design of the large pebble bed high-temperature reactor and have verified the mechanics of the pebble bed core on original scale.

Thus BBC/HRB have the know-how required for designing, licensing, constructing high-temperature reactors constructioe th r first-of-it.e Fo th f no s onle th y 0 require10 R HT -d kinan d0 plant50 R HT s ment remainin researca s i g d developmenan h proD tR& - gra confiro t m e designth m . These R&D activities conducted in parallel have muca o beeht largeR n LW performe e r th extent r fo d , since they have been continue substantialld an d y sponsored form public funds even afte construce th r - tion of several LWR plants. e thresholth t a f commercialis o di R HT Thue th s - zation and is ready to be practically used by public utilities and other potential customers on the thermal energy market.

2. HTR LINE WITH ITS MAIN CORNERSTONES HTR 100 0 (FIG50 R AN.) 3 HT D Based on the constructed AYR and THTR 300 plants, BBC/HRB develope planR dHT t concept powee th rn si levels between 100 (HTR 100) and 550 MW (HTR 500). These usee plantr b fo dn ca s - electric power generation (also in arid areas by using dry cooling towers) combined generation of electricity and process steam up to 530 °C by co-generation of heat and power combined generatio electricitf no districd an y t heat. economin a r Fo c utilizatio domestif no c coal the HTR represents an environmentally favorable and resource-saving solution, whe operatet ni s combined with conventional gasification procedures and especially in direct application of the nuclear process heae appli higt Th ts temperaturea . ga h- °C 0 95 f o s cation of the HTR development which is being persued withi nucleae nth r process heat project (PNP) under substantial sponsorship from the Federal Ministry for Research and Technology and the Northrhine Westphalia Ministr Economicr fo y Technologyd san .

238 1200

Power MWe - tr

550

300

200

100-

FIG. Powe3 , liner R Spectru.HT e th f mo

The generation of SNG (substitute natural gas) offers the possibilty to make use of the available domestic natural gas piping systems and markets for generateG SN d from German lignit hard ean d coalt .A same th e timeG , SN import reducee e b th n y ca sdb quantity produced, which result noticeabla n i s e balancreliee th f paymentsf o eo . Another possibili- productioe th s ti y f liquino d hydrocarbons which are capable to replace mineral oil in case of necessity. Also for this long-term development objective, the direct application of the high gas temperatures up to 950 °C e.g. for coal gasification or for the application of the gas turbine, the main components of the nuclear heat generating system of the HTR 500 and the HTR 100 can be essentially adopted. However, further developmen requires heat-exchange i t th r fo d - ing components, hot gas valving etc. until a commercial application wil possiblee lb . This development wil e successfulllb thif o ysd completeen e th y b d century provided that the required R&D support will be provided from public funds also in future. presene Baseth n o dt energy situation, applica- coar fo l R gasificatio- HT ex e e tioth b f o no t s ni pecte befort no dbeginnine eth nexe th t f centuryo g .

239 3. THE HTR 500 POWER PLANT 3.1 The_Technical_Concegt

In Fig. 4 the general arrangement of the power plant is compared with the THTR 300 on identical scale. It is optically quite evident that the plant dimensions are practically the same for a power out- whict almoss pu hha t been doubled.

F«MI «Mtar tw* »nd start-up

HT0 R50

FIG. 4. Comparison of dimensions of THTR 300-HTR 500.

powe0 50 rR planHT centee e Inth tth complef ro x the reactor containment building is situated housing the prestressed concrete reactor vessel including the primary system shutdowe ,th n facilities, parts decae ofth y heat removal syste othed an m r safety- relevant components. In addition the reactor containment building protectioacta s sa n agains- tex ternal impacts. All the component parts carrying activity are arranged inside the reactor containment building except the fuel element storage, which is, however, also equipped wit protectioha n agains- tex ternal impact. The reactor service building on the reactoe th e sid f on eo r containment buildine th d an g turbine building on the other side form a line of buildings to which the electrical equipment building is arrange oppositn a n i d e position.

240 Fig. 5 shows a vertical section of the prestressed concrete reactor vessel witinternalss hit . The fuel elements contain low enriched uranium and pass through the reactor core only once (OTTO fueling) . The spent fuel elements are discharged from the core through three fuel element discharge pipes with- t interruptioou powef no r operation spene .Th t fuel elements are directly discharged into casks and transported to the storage. No special conditioning for spent fuel is necessary for longterm interim storagr finafo ld ean storage .

026500 »ncorerod

Main circulator Reflector rod Main steam /tine Auxiliary heat exchanger

Codling water

Auxiliary - circulator

discharge pipe

FIG. S. HTR 500 — Reactor pressure vessel with internals.

The waste management concep adaptes e ti th o t d special characteristic lineR year0 1 HT .e A s th f so interim storage of the spent fuel elements in fuel storage cask s intendei s d followe y transporb d o t a nuclear waste repository for ultimate disposal. Due to the limited quantity of spent fuel elements, fueR HT l closin e suitablo cyclth n f s o i eg e solution, so that - as for other prototype nuclear power stations - direct ultimate disposal is envisaged in e lonth g run. This concep THTd tan RcorrespondR AV e th o t s 300 waste management concept s license,a e th y b d authorities.

241 The reactor pressure vessel is designed as a large cavity prestressed concrete reactor vessel (PCRV) as in the THTR 300. The complete primary system is integrated in the PCRV. s equippei 0 50 dThR witeHT stea6 h m generators of helicoil design operatin countercurrentn i g . Each steam generator is associated with a circulator. The circulator e THTsame desig th th s thaa Re f s to i n circulator, it is however, equipped with active magnetic bearings and arranged in vertical position. The helium coolant flows downward through the reactor core, then upward throug steae th h m generators to the circulators from where it is recirculated to the cold gas plenum, cooling on its way the outer surfaces of the primary system. Because of the availability of the separate two-loop decay heat removal system the main cooling system consisting of 6 steam generator/circulator units has no safety function. Therefore the main coolin gpurela systef o ys i mconventiona l design outside the reactor containment building, i.e. designes i i te usua th lo t d standards applieo t d modern conventional power plants.

1 Reactor core Coolin6 g water system 2 Steam/feedwater circuit PCR7 V safety valve Mai3 n heat sink 8 Reactor containment pressure relief system Deca4 y heat removal system Exhaus9 syster ai t m Line5 r cooling system 10 Ventilation stack D Components supplie emergency db y power

FIG. 6. HTR 500 — Flow diagram.

242 etY_Concegt__f §a 2 3. f or_Accident_Control__(Fi2_1_6 )_

The accident control concept of the HTR 500 is based on the following principles: Multi-barrier principle for activity enclosure Safety systems for reactor shutdown and decay heat removal HTR-inherene th Usf eo t safety characteristics for taking manual measures in the event of failur activf eo e safety systems.

Multi-Barrier Principle

The following barrier providee ar s d against activity release: Coated particles (TRISO) Fuel elements Prestressed concrete reactor vessel with leak-tight liner Reactor containment building The design of the fuel element and especially the build-up of the coated particles ensures a low releas activitf o e y during normal operatios a d nan a resul accidentsf o t .e even Eveth hypothetif n to i n - cal accident n increasesa d releas f activito e y from e fueth l element n onlsca y e operatinoccuth f i r g temperatures are substantially exceeded. Such high temperatures would only be reached as a result of fail- decal al yf o hea e tur removal system onld an sy several hours afte occurrance th r e accidentth f eo . There- fore sufficient tim availabls ei o takt e e manual measure interrupo t s accidene tth t sequence. The reactor pressure vesse burst-safa s i l e prestressed concrete reactor vessel equipped with an inner liner line,a r cooling systethermaa d an m l insulation. Large-scale failure of the primary system containmen rulee b n dtca out. The greacter part of the vessel closures is vessele th heap f to arrangeo d . e th Penetration n i d s for the helium purification system and the fuel circulating system pass throug bottoe th h m e platth f eo prestressed concrete reactor vessel. Vessel closures and penetrations are so designed that large leakages from the primary system can be safely ruled out, e.g. by installation of redundant structural elements. The leakages to be assumed are improbalbe and limited to a maximum of 33 cm2.

243 e reactoTh r containment buildin designes i g o t d withstand earthquakes, external shock wav aird ean - plane crash. Primary gas leakages due to operation or acci- dento leat kp su size approximatelf so e ar 2 cm 2 y vented to the environment under controlled conditions through filter systems and the stack. maximue Th m leak cross sectioPCRe s th Vha f no y beestructurab 2 n cm limite 3 3 lo t dmeasure n i s sa the THTR. As a result of this design, there can only be a slow leakage of the coolant; the time for pressure equalizatio approximatels i n minutes0 9 y n .I this case, the coolant gas is passed straight into the stack unti depressurizatioe th l n reactoe systeth f o mr containment building closes again. As a result of such low coolant gas activity, the potential radiation exposure of the environment of the nuclear power plant originating from the radioactive substances released less i witvene sr th hthatai n 1 mrem/a wholebody dose. The environmental load in the event of accidents is about 10 mrem wholebody dose 3.3

Reactor scram following accident automaticalls i s y effected by the reflector rods which drop into bores in the side reflector by gravity. In the relatively rare event f long-terso m shutdown onle incorth y e rods are inserted directly intpebble o th core d ebe . Long- term shutdown is exclusively actuated by manual release being generally required only after 24 hours in order to reac cole th hd subcritical state. The decay heat removal concept (Fig. 7) is characterized by maie th n f o hea e tus removae th l system a separate and redundant decay heat removal system (two loops) with particular circulators and heat exchangers in the primary system utilizatio naturaf no l convectio decar fo n y heat removal in the primary system - possibilit prolongea f o y d failurdecae th yf eo heat removal system and subsequent resumption of operation following manual measures integration of the liner cooling of the pre- tresses d concrete reactor vessel intdecae oth y heat removal concept without activity release in the event of total failure of the decay heat removal systems listed above

244 T_yes_ Decay Heat Removal Main Loop Cooling by Main Loop Cooling (MLC)

Aux H«at Removal Decay Heat Removal System (2x100%) _yes_ x SysAu t2 r o b y1 CACS(1) CACS(2) Feedwater Circulator* Decay Heat Removal by Natural Convection to Aux Heat Exchanger (LOC) Repair and Emergency Measure instal- re o st l Resumptio f Decano y Aux or Main Cooling Heat Removal System withi hour0 n1 s

Decay Heat Removal I——V£S_ Liner Cooling System by Natural Convection to Liner Cooling System (LCS)

Repair and Emergency Measure re-instalo st l ves I the Liner Cooling System withi hour0 n1 s

Hypothetical Range « 10 6 /a

FIG. 7. HTR 500 — Multiple measures for accident control.

hypotheticae th n i - l cas f totaeo l failurl al f o e decay heat removal possibilitie onl) /a y ~ 10 < ( s limited radioactivity release to the environment. 3.4 Accident_Behaviour_of_the_HTR_500

Accident analyses s performea THTe 0 th 30 R r fo d wel s risa l k analyses have shown thae desigth t n criterieffectivenese th d an ae safet th f yo s systems are mainly determined by the following accident categories : Failure of individual decay heat removal systems or loops, i.e., reduced decay heat removal capacity. - Reactivity accidents, i.e., higher core power to be removed. - Depressurization accidents, i.e., reduced primary circuit heat removal capability. Tube ruptursteae th mn i egenerator , i.e., penetratio f wateno r intprimare oth y circuit, followed by a rise in primary gas pressure an increased corrosion of the graphite internals.

245 Moreover, the control of failures of total systems was investigated: failure of the scram system failure decath yf o ehea t removal system. Asummara s y resul f thesto e analyses Fig8 . shows the time curves of the maximum fuel element temperature e accidentth r fo s s listed above compareo t d e load failurth an d e e fuellimitth f .o s

Fuel Temp

3500 Graphite Sublimation

3000--

2500-•

2000-—— Particle Failure

100% Cs'Diffusio n= MaiMLnC j_oo p Cooling 15OO <001°'o CACS (21 = Core Auxiliary Cooling System (Z Systems) Design-Temperature = torCAC ) e(1 S Auxiliary Coo g Systelin 1 Systemm{ ) ATMS = Anticipated Transient k/ithout Scram Normal Operation 1000 LOC = J.OSS of Circulators = n Combinatioi CAC ) ) (1 (1 S n with Depressurization (Depress.) LCS = Uner Cooling System 500

100 Time [h]

FIG. 8. HTR 500 — High temperature resistance fuee oth f l elements.

evene Ith n f desigto n basis accidents, i.e. accidents whose controllability has to be verified in the licensing procedure, forced-circulation cooling of the cor maintaineds i e sucn .I h event fuee th sl element temperature e reducear s o temperaturet d s with C ° belo 0 - 50 w in a few hours, so that even most improbable accident- failure combinations involving water or air ingress can e controlleb d withou y problemstan . (BeloC w° som0 50 e the reaction rates of steam or oxygen with the graphite e fueoth fl element e negligiblar s y small). In addition, Fig. 8 shows that even in the event of improbable accident sequences such as total failure decae oth f y heat removal system r failurso e th f eo scram system (ATWS accidents) the fuel element temperature will rise for a short period, but will remain clearly belo e desigth w n temperature permitte conr fo d -

246 tinuous operation. This ensures that an increased release of activity from the fuel elements will not occur. HTR-Inherene th f o e 3.us 5t Safety Characteristics

The inherent safety characteristics of the HTR, i.e. the low power density, the high failure temperature coatee ofth d particle e integratioth d an s n int preoa - stressed concrete reactor pressure vesse usee n ar li d the design. The prestressed concrete reactor vessel maintains s geometricait l n importanford a act an ms sa t barrier against fission product release, since part of the fission product depositee ar s d thuan ds retained within the vessel. The permissible failure period of the decay heat removal . h system 0 1 s i s Accessibilit e externath f o y l componente th f so decay heat removal syste ensuro provides t i m s a e o s d that the required measures can be performed with a sufficient reliability withi periona hours0 1 f o d. s beeIha t n demonstrate greaa n i dt numbef o r analyse e accidenth n so t behaviou plantR HT f sro that there is a possibility of an even further reduction of e anywa th accidenw lo y e line th te HTR rf th risi , f o k cooling prestressee systeth f o m d concrete reactor vessel, operating in normal conditions, is integrated decae ith n y heat removal conceptvera s yi simplt .I e and therefore very reliable system for decay heat removal which can rapidly be taken back into operation by simple measures such as mobile aggregates. Therefore it has been designed so as to ensure tha e eventa long-ter th evef o tn i n m total failure decae oth f y heat removal loops wite reactoth h r pressurize activito n d releaseds i y pressure .Th e e primarth ris n i e y system resulting froe temth m- perature increas limites i e y overflob d f heliuwo m coolant inte storagoth e heliue tankth f mso purification system. Core temperatures remain below 120. °C 0 In the extreme event of additional failure of the liner cooling syste e inerth m t behavioue th f ro system provides sufficient time to take measures for restartin e lineth g r cooling syste y emergencb m y feed.

Analyses performed by the Institute for Nuclear Safety Research in the Nuclear Research Center Juelich have shown that even in such hypothetical events there will be no immediate effects and the collective risk of late radiation damag vers i e y low.

247 3.6 Cost and_Economics

An HTR with a power level of 550 MW has the advantage of a better correspondance with the increase in power demand and thus lower investments to be made in advance as compared to the large LWRs. Additional advantages result with regard to power reserve and mains loading. These advantages may be decisive in case of otherwise almost identical electricity- generating costs. The consequent utilizatio HTR-specifif no c characteristic optimization a wels sa s a l e th f no design of components and circuits leads to the result that the electricity generating costs of an HTR 550 are competitive to those of the large LWRs. 3.7 §tate_of_the_HTR_500_Project

In 198 usere 3th s group Arbeitsgemeinschaft Hochtemperaturreactor (AHR), representing 16 power industry companies, place orden a d r with BBC/HRB to perform conceptual design studies on the HTR 500 oprivatna e business basis. These studies have been successfully completed. AHR has verified the results of the HTR 500 conceptual design project thin .I s evaluatioe th n positive statements on licensability technical feasability - economics and competitiveness with the large 1300 MW convoy pressurized water reactor have been confirmed also frousere th m s side. This evaluatio summara documentes n i nwa R y AH repor y b d t in July 198 distributed 4an utilite th o t dy companies. In addition, the technical and safety concept of the HTR-500 has been confirmed by an experts committee invoked by the Federal Ministry of the Interior, which was composed of members of the Reactor Safety Commission, the technical supervisory association (TÜV) and the licensing authorities. Base thesn o d e positive results Germae ,th n companies operating nuclear power plants took a fundamental decisio projec0 50 continuo t nR tHT e designeth . According to the state of the negotiations with the users group Hochtemperaturreaktor Gesellschaft (HRG) e expecteb n ica t d that BBC/HRB wil awardee b l e th d two years' contract for the design of the HTR 500 neae ith nr future. This period include nucleae th s r licensing procedure for the concept of the HTR 500.

248 Immediately afterwards, Part B of the design contract placee b o it sd which cover complete th s e nuclear licensing procedure so as to rule out delays during contruction resulting froproceduree th m . Because of the possibility of future application of the HTR for process steam and process heat, the Union of Chemical Industry and Ruhrkohle AG will participate in the project.

4. STRATEGY OF HTR COMMERCIALIZATION mentionee Baseth n o d d fact e followinth s g strateg e commercializatioth f s o i y R HT e th f no envisaged: - Desig constructiod nan 0 50 firse th R tHT f no plant, i.e plan.a electricitr tfo y generation witpossibilite th h f extractinyo g process steam and district heat in the Federal Republic of Germany powee .Th r outpu meetW f abouM to 0 s55 t the modified demand of the power plant market, whic r reasonfo h f grio s d size, scop f financeo - ing and the lower increase in electricity demand internationally shows a trend towards power units . MW 0 60 - 0 e rang30 i th nf o e The possible start of construction of the HTR 500 will be 1991. Design and erection of the HTR 500 are financed on a commercial basis. The design work is supported by a R+D program sponsored from public funds. The realization of the first HTR 500 will render competitiv0 50 R n electricitHT a e s ea th d an y process steam generating plant. Further contracts acceptee b n worldwida ca n o d e scale froe th m early 9Oies onward. greaa 0 t50 plane th parR tHT f to come Foth r - ponents can be fabricated in the Third World countries by their domestic industry, including the erection of a prestressed concrete reactor vessel. Joint ventures have recently become common practice. Potential customers are Third World countries havin correspondina g g electricity demand, such as China, Taiwan, Indonesia, Turkey othersd ,an . e smal Th designeR HT l powea o t dr outpuf o t s envisagei W M r 0 speciafo 10 d l application i n industry (steam and electricity as combined pro- alsn constructee ca ob W M 0 10 R d ductsHT e Th ). as a 200 MW twin plant which is economically competitive with fossil-fueled power plants and hence competitive with conventional plantf o s identical size without involving the problem of air pollution.

249 For this purpose an industry consortium HTR 100 was foundecompaniee th y b d s BBC, HRB, Deutsche Babcock, Mannesmann Anlagenbau, Straba Innod an g - tec. It is the objective of this consortium to establish a detailed design of a competitive plan simplificationy tb , standardization, series production, shop fabrication and development of a transport and assembly system for prefabricated plant components. Also for this project a supporting R&D program is conducted with the sponsorship from public funds. Currently a users consortium for the HTR 100 is being established with the objective of award- ing a contract for the design of the first-of-its- kind e Federaplanth n i tl Republi f Germanyo c ; construction of the plant will be possible in the early 90s. For Third World countries havin a loweg r electricity demand and few fossil resources, the HTR 100 represents an excellent possibility for generating electric power and heat. Extensive participation of the domestic industry is ensured. r opinioIou n n thergreaa s i et numbe countrief o r s that wil potentiae 0 frob l 10 m R HT le userth f o s the early 90s onwards. The heatinW smalM 0 1 l g reactor - offerex n a s cellent opportunity for providing densely, populated areas with space heat which eliminates their problem r pollutionai f o s . Swiss industry envisages the construction of a prototype plant in Switzerland in the next years. A successful f theso e eus plant possibls i s mann i e y countries of the world. From the mid-nineties onward also the constructio e generatiof plantth o n r fo s f nucleano r process heat up to 950 °C will be possible. This date will, however, finally be determined by the market. Based on the present energy situation, this application is to be expected not before the beginnin e nexth t f o gcentury .

250 SAFETY ASPECTS (Session D) SAFETY CHARACTERISTICS OF CURRENT HTR

W. KROGER Institu r Nulearfü t e Sicherheitsforschung, Kernforschungsanlage Jülich GmbH, Jülich, Federal Republi f Germanco y

Abstract

The paper contains a description of HTR-safety features and results of safety calculation e HTR-500th 0 MW(e r 55 fo s)a , pebbl HTGRd be e A . f advanceo t se d safety criteri s proposei a d discussedan d .

The objectives of statutory regulations in W. Germany are . to avoid accidents by means of high technical and human quality controo t . l accident meany sb reliablf so e engineered safety features and . to mitigate the consequences of severe accidents by technica administrativd lan e measures, e.g. emergency planning. Limits for radiation exposure during normal operation and incident design-relevanr welfo s sa s la t accidents are set up by the Radiation Protection Ordinance (3wholm O re mrem/5 e d bodaan y dos referencf eo e person at most adverse location outsid plant)e eth r sever.Fo e accidents radiological establishedt limitno e sar ; however, there are reference values for protective actions planne advancn i d remotd ean e siting criteria mainly to facilitate evacuation.

On the basis of these regulations a set of design relevant accidents and safety requirements have been developed for light water reactors in a far reaching de- terministic manner. The result is . a high safety standard assured by a highly active complex safety system with thre r foueo r redundant

253 trains and no need for operator actions within the first 3O minuts, . a small possibility for (beyond design) severe acci- "*" *• 1 ' whicr dentfo ht sbu catastrophi c activity, releases and consequences cannot be excluded, and smala . l risk considered justifiabl Supreme th y eb e Court hardlt ,bu y accepte s structurit e n th i d y eb public. For non-standard reactor systems, like HTR o firmln , y established design rules and standards exist. Liter- ally fulfillment of LWR requirements is possible, as e licensinshowth n i n g procedur THTR-300e th f o e t ,bu it leads to an inpart overloaded safety system if the passive safety feature differene ar s morr to e favourable, respectively.

Therefore effor bees tha n mad recenn i e conceptR tHT s to develop a more specific safety concept based on the general sense of legal requirements and, according to their nature, making use of Probabilistic Safety Assess- ment. This concept (see Table I) has been applied to the HTR-500 and HTR-Modul under development and judged positively by a group of experts established by the upper licensing authority /1/.

TABLE I Proposed Concep r Frequency-Orientefo t d Safety Requirement Advancer fo s d Reactors

Radiological Limits Event Classification, Explanation of Frequencies (Whole-Body/Thyroid Dose Risk) Accidents relevant for design > 310-Vyx r Event not to be excluded during lifetime 30/90 mrem/yr (§45 Strl SchV) plante th f o 20 ..1 .3 x lO'Vy r Event not to be excluded during lifetime <5/l5 rem/event of several plants 10-"... lO'Vyr expectee b o Event t dno t during lifetime 5/15 rem/event (§28.3 Strl SchV) of several plants

Severe accidents <10-5...l

"Strl SchV: Strahlenschutzverordnun gradiatio= n protection ordinance.

254 Very promising passive safety feature R (seHT ef so Table II) on one hand and a need for urban siting on the other encouraged us to think informally about more restrictive safety requirements. In the light of Chernobyl-desaster thebecomy yma e more importans ta generaa l proposa r advancelfo d safety criteria (ASC) and may help to overcome public resistance against nuclear power. In any way, they should be used as starting point for a discussion on an international level aiming at common or comparable safety criteria for super-safe advanced reactors.

TABLI EI Important Safety Characteristic f HTRo s s Relativo et Standard LWRs (Values for HTR-500)

• Larger heat capacity (~103 ton f graphite)so , much lower power density (6 MW/m3) -> Slow reaction on loss of cooling (500°C heatup after 5 to 10 h) • Distinct negative temperature coefficien f reactivito t y -> Fast shutdown system not necessary • Ceramic core, ceramic fuel coating with high retention capacity even under core heatup conditions -» No fuel melt, fission product release small and delayed • Single-phase inert coolant (helium), lower pressure level (-50 bar) -> Loss-of-coolant accident with fewer problems, nee r (forceddfo ) convection potentiaw lo , r fo l mechanical destruction • Potential for massive graphite corrosion if T> 500°C and O2 > 1 ton -» Air/water ingress accidents significant • Coolant gas activity insignificant, plateout and dustborne activity relevant

The basic idea for the proposed ASC (Table III) is to limit the activity release resulting from severe accident y thasucn wa i s t a h emergency measure neet no d o d s tconsideree ob protectioe th publie r th fo d d f ncan o the financial damage can be coped with by the society. Emergency measures include . acute measures like evacuation, distribution of stable iodine tablet earld san y relocatios na well as

255 TABLE III Advanced Safety Criteria ( ASC ) propose Institute th y db e

limio t s ti activitThm eai y releas severf eo e acciden thay sucn i t wa t hemergenca y measures do not need to be considered ( no evacuation, relocation, decontamination ) and the financial damag copee b n deca wit h

The way is to establish short-term and long- term dose limits for individuals at worst location, rangef so 1 to25rem for 7 days exposure eary 0 s3 exposurr fo m re e0 25 1o 0t

Reactor with passive inherent safety characteristics is most promising for fulfilment

. late measures like late relocatio ared - nan ade contamination . r thiFo s purpos dosea e limi r short-terfo t days7 ( m ) and long-term exposur afte) a 0 r (3 esever e accidents is proposed for individuals at worst location. Reason- able values have been derived from intervention levels for protective actions showing dose rangef so 1 to 25 rem for short-term exposure and r long-terfo m re m 0 exposure25 o t 1O .

For the conversion of release into doses calculation principles should be applied which have originally only been vali design-relevanr fo d t accidentd an / s/2 which are slightly reduced in their degree of conser- vatism for this purpose.

Severe accidents include all events and event sequences of such a low probability that the plants do not designee b nee o t d d against them;^ They normally domi- nate the risk and need to be identified by comprehensive probabilistic safety analysis (PSA). Frequencies e rangith n f loweeo r tha" 10 n 1O-8 per reactor-

256 year are proposed as cut-off line. Vulnerability in relatio acto n t sabotagf so catastrophid ean c rare events should be as small as possible. Assuming 2,5 rem as dose limit for 7 d and 25 rem as an even more stringent limit value for 30 a exposure time aproposese welth s la d conservative pricipler sfo back-calculation the tolerable release of the representative nuclido t rangO e e2 th Cs-13f n eo i s i 7 10OO Ci depending on technical parameters like emission heigh timed tan . t intenno dictato Wo t ded technicae eth l measurey sb whic fulfillede b hn knoe thesca w wt C e.Bu AS fro m PSA experience and risk reduction studies for miti- gation features that a reactor design with passive inherent safety characteristic mose th t s promisinsi g way: The reactor may not loose its retention capability even under total los coolinf so g conditione sdu to physical reasons attempe ;th excludo t e practically loss of cooling conditions by providing an additional active safety system is misleading. The course of severe accidents must be slow with timmitigatinr efo g countermeasuree th r fo d san reversal of wrong human actions.

questioe Th whethes ni safete rth y characteristif co current HTR concepts of small (HTR-Modul of the KWU- group, HTR-10 BBC-groupe th f 0o mediur )o m power (HTR-5O BBC-groupf Oo appropriate )ar meeo et t these ASC.

As lon massivs a g e corrosion excludeattace b n kca d temperature induced coated particle failure is the dominant release mechanis fissior fo m n products. THERMIX core heatup calculations show (Fig. 1) that for small HTR the maximum temperatures for fuel element beloe s ar slightlr wo range th f increaseen o i y d

257 depressuri2 ed total particle L juu failure /

2000- J(HTR-100)|/y ,—— ii V ——— —— H -— increasing / permeability 1500- t X1 / pressur.zed — — - —— design value 1000-

500- HTR-Modul (THTR-300) HTR-500 (3MW/m3) (6MW/m3) o-l 200 0 50 TOOO 1250 Power lMW,h]

Fig: Maximu1 . m temperature f fueso l elementr fo s core heatup accidents in pebble-bed HTGR

Reactor protection building

ventilation system

Relief system

Mam cooling system Nuclear service water system (2x)

(T) Afterh«at removal syste) x m(2

) (2 Blower cooling

) (5 Liner cooling system (2x| with emergency feed

(Î) Steam generator dumping system

^ (f Emergency power

Fig. 2; Cooling system and confinement of HTR-5OO

258 permeability, whereas they reac range hth totaf eo l particle failure in case of the depressurized uncooled HTR-5OO. Therefore wan,I limio tt subsequene tth t discussion examplmainle th o .HTR-50Ot f yeo originae .Th l plant desig shows ni Fign n i meanwhil, 2 . s beeha nt ei further developed by the vendor.

HTR-50e Th W (electric0M cora sphericaf s eo )ha l fuel elements cooled by helium in a downward flow. A prestressed-concrete reactor vessel (PCRV) with a steel liner contains all the components of the primary circuit thermaa ; l insulatio redundana d nan t liner cooling system (2 x 10O%) protect the vessel against excessively high temperatures. Apart from the main cooling system, a two-train auxiliary cooling system is availabl r afterheafo e t removal startes i t .I d automatically and one train is sufficient even with an unpressurized reactor. On the primary side, the afterhea n als tca removee o b y naturadb l convectio fult a n l system pressure. This require sspeciaa l operator action from the control room after half an hour. The final activity barrier to the environment is a vented confinement that filters small leaks froprimare mth y

TABLV I E Core Heatup Release Categories for HTR-500*

Cumulative Fractio f Corno e Inventory Released to Atmosphere

Frequency Time Kr-Xe Iodine Cesium Sr-Ba Te-Sb Category 1/yr (h) (1.3x 10«) (1.7x 10") ( 910*x ) (l.8x 10») (4.6 x 10')

Kl core hcalup Oto 35 1.4 x 1C'2 2.4x IO0 3.3 x IG'3 3. 9lu'x J 30 1 x 4 2. PC open 10-' 0 20 o t 0 2.2 x lO'2 ° 10 2. 4x 3.3x 1' 0• 4 x 10 3 2.4 x 10 ' Failure of LCS (LoPS- 100%) 2 >200 2. 2x IO- 2 2.4 x 10-' -io- IO"x 3 8 - 2.4 x 10 ' RPB stack cor2 K e heatup 3x 10-' 5 3 o t 0 1. 4x IO' 2 2.4 x 10° 3.3 x 13 0 0 1 •' x 9 3. 2.4 x 10 ' opeC P n (LoP S- 100 %) >35 2.2x IO-2 2. 410-x ' 3.3x IO-3 IO'x 4 ' 2.4 x 10 3 RPB stack K3 core heatup 10-' 0 to 1- 2 10 x 3 io- io-' I.I x 10 s 10 " PC open Oto 35 1. 4x IQ' 2 10" 1.5 x 10 s 3. 19x 0 ' 10 4 (LoPS 90%) RPB filter stack >35 2.2 x IO-2 io- l.5x 10 ' s 3.9 0 1 x to- K4 core healup io- shorC P t term open Oto 12 4x 10' 2x 10' --- 2x 10 " (LMCS 85%) — - RPB filler stack

Noteprimar= C P : y reactocircuit= B RP r, protection building (confinement) contributio= ) ( , o t frequencyn , LoPS• loss of preferred power supply, and LMCS - loss of main loop cooling system. •Initial inventories of grouped nuclides are in parenthesis.

259 Los f preferreso d Loss of main ^ power supply (LoPS) cooling system) (LMCS) 4 x 10~2/yr 0.4/yr

Start dieser lo restoration of grid supply <5 h

/ * Removat f o l afterhea/ as t Yes designed; No

•Restoration of 1 1 Core heatup grid supply only Core heatup reachin point gse f to reaching set point of PCRV-RV afteh 5 - r PCRV-RV h afte 5 - r and failure to close. and failur closeo et , partial depressurization partial depressurization

Restoration N Yes of energy supply S10 h

LCS emergency LCS emergency feeding s 10 h feedinh 0 1 gs

<4 —— ———— —— Yes Yes 1 1 CORE HEATUP CORE HEATUP PC OPEN PC SHORT TERM RPB-FILTER-STACK OPEN RPB-FILTER-STACK

K1 2x1(T7/yr K3 9 x lO'Vyr 10-'/yx 2 K2r x 1Q- 2 4 7K /yr RP8 = Reactor protection building RV = Relief valve AHRS = Afterheat removal system Mai= S n coolinMC g system LCS = Liner cooling system PC = Primary circuit

Fig ; Dominan3 . t event sequence r corfo se heatup categories of HTR-5OO

260 circuit and releases larger leaks up to a maximum rate of 11,5 kg/s unfiltered intenvironmente oth botn ,i h cases via the stack.

To establish the accident topology, parallel to the planning work, more than three smalyeara o lsag effort PSA /3/ has been carried out by the Institute. Dominant initiating events are not so much leaks in the primary circui so-calles ta d transient events release .Th e category which dominates the risk and results in the highest release figures is initiated by loss of preferred power, followe reactoy db r scram, failurf eo diesels to start and non-restoration of grid supply within five hours for the afterheat-removal-systems, non-restoration of energy supply for the liner cooling system and no emergency feeding within 10 hours (Fig. 3).

The slow temperature rise is illustrated by Fig. 4. A periot leas a repaie remainh f th 5 tdo f r ro sfo afterheat removal system or restart of power supply and at least 10 h to restore liner cooling.

Only a few percent by volume of the core reach temperature maximua f so 2250°Cf mo . Large section- sre main below 1600°C and do not contribute to temperature- induced fission product release which explains even for this low frequency event sequence (é 2-1O /reactor- year) relativel releasw lo y e values.

Release intenvironmene oth t doe begit sno n until afte wheh response 5 nth r~ e pressur PCRe - th re V f eo lief valve bees sha n reached thef .I y fai closo lt e the first phase of release is finished after 35 h when the depressurization is completed. An additional 'activity thrust" had to be expected in the long term thermacausee th y db l degradatio unprotectef no d concrete of the PCRV in the top area.

261 100 200h

HTR-500

500 1200 1600 2500 Temperature [°C] Fig: 4 Percentag. volumactive y eb th f eo core region versus temperature at different time for core heatup

core Inth e tota f o inventor % l1 cesiumf yo , this means 3-10 Ci of Cs-137, would be released unfiltered via stack. In comparison to tolerable release values derived from the proposed ASC they are too high by one or two orders of magnitude. But nevertheless, it seems to be feasible to meet these stringent requirements even with the HTR-50O if we take new results of safety research and design options for passive systems into account.

The retention of_fission_products by absorption in the reflectore graphitth core d th ean f eo s have been treated inadequately in the past, that means in the small-effort PSA. Taking absorption of metallics beyond 1500 C into account by extrapolating available data accordin knowo gt n physical laws, this would a lea o t d reductio cesiuf no a mo t releas p u factoa 5 y eb f ro potentia morf lo e thaExperimenta. 1O n l worbees kha n initiate o improvt d date matrie th a th d bas r an x efo reflector graphite and to confirm that absorption pro- describee b ces n sca steady-staty db e assumptions.

More sophisticated thermohydraulic calculations have shown that the afterheat can be dissipated into the massive PCRV-structure (Fig. 5, maximum core temperature below 160O°C after hal yearfa ) without jeopard- izing its consistancy.

262 300O

Centre Outersurlsce of Core Insulation Llnor 0.25 m Concrete 0.7Sm 1.80m 3.80m

500O 1O.OOO 15.OOO 20.0OO Time (hi

: FigTemperature5 . reactof so r regions versus timr fo e los f corso d line ean r coolin HTR-5On i g O

The temperatures of the liner including its insulation keep below 12OO° Cheatue th eve d vaporizatiof i an np n of the water is neglected conservatively. In the past we have assumed that the liner would fail at temperatures of 1000°C, but current experiments have shown more favourable values muce b hy ,highema r than 12O; OC the steam produced is externally released through the pores and cracks of the concrete. If this can be confirmed by more representative experiments with segment t froPCRVe scu th m ^ thermal degradatioe th f no concrete (beginning after more thao weake n non s )ha longer to be assumed. The second phase of release would be eliminated leading quantitativel reductioa o t y n of cesium release by a factor of 3. Even if this failure mode would not be impossible the amount of water released from heated up concrete and enterin core expectes th gi e muce b ho t dles s than assumed in the past. The maximum hole in the PCRV is 2 sizn i e m c (relie 3 o3 f f e safettraith f nyo valve); it limits the mass flow into the reactor building to values which allow filtering of the exhaust. According to recent studies metal fibre available filterar / s/5 e which are characterized by their temperature resistance pressurd an C O e 3O differenc o upt e resistanco t p eu

263 1 bar; decontaminatio e reacheb n nca d factor 0 1 f so for cesium aerosols .passivelA y arrangemen thesf to e filters withi reactoe nth r buildin possibles i g e th n I . case of a rapid depressurization these filters made from metal overloadee fibreb n sca factoa y b df ro compares 1 0a normae th o lt d design.

In conclusion, it is clearly possible to meet non- catastrophic release requirements, technically formula- proposee th tey b d d ASC y smalb , d potentiall lan y even by medium HTR. Although some restrictionse b hav o t e made: Mor- e comprehensive safety studies includine th g whole spectru f initiatinmo g event evend san t sequences, including change reactivitn i s r o y airwated an - r ingress, hav confiro et e th m result small-efforf so t assessments trende ;th s derived from current research work e b hav o t e verified.

- The proof of safety for passive components and physical features, which largely determine the advanced safety standard, must be further raised in quality. It mus showe tb n tha catastrophita c failurf eo the pressure vessel (PCRV may be more advantageous the steel vessels), large closure piper so n ca s be exclude doer o dt lea higheo sno t d r consequences cause massivy b d e graphite corrosion.

- Safety concepts which do not only rely on the retention capability of the fuel elements, but on controlled releas filtersa evi e ,th likr fo e modified HTR-5OO, hav mako et e sure thae tth filters cannot be bypassed under severe accident conditions.

These reservations should motivate further work and cooperation, but they do not doubt the attractiveness of HTR from the safety point of view.

264 REFERENCES

/1/ Kroger Weiters, ,W. , J.P., "Experience th n i e UsProbabilistif eo c Safety Analysie th r fo s Developmen Safetf to y Concept Commerciar sfo l High-Temperature Reactors", Nuclear Technology, Vol. 74 (1986) / Empfehlun/2 RSK/SSKn gvo : "Störfallberechnungs- grundlagen für die Leitlinien des BMI zur Beurteilung der Auslegung von KKW mit DWR gemäß Abs8 2 StrSchV§3 . , Bundesanzeiger , 24535 ,a (1983) / KFA-ISF/3 : "Zum Störfallverhalte HTR-50s nde 0- Eine Trendanalyse", Jül-Spez-220, Kernforschungs- anlage Jülich (1983) /4/ Wilhelm, J., "Development and Use of Vent Air Filter Nuclear sfo r Power Plants" ConcludinS ,PN g Colloquium, Karlsruhe (1986)

265 THE SAFETY CONCEP MODULAE TH F TR O RHT

G.H. LOHNERT Internationale Atomreaktorbau GmbH, Bergisch Gladbach, Federal Republi f Germanco y

Abstract

e papeTh r contain mose th st important design feature modulae th f o sr HTGR with pebbl descriptioa fued d be elan modulaf no r HTGR safety feature d safetan s y concepts.

Components of the nuclear steam supply system are described together with their impact on the plant's safety behaviour.

1 INTRODUCTION

modulae Th poweR rHT r plan universalla s ti y applicable energy source for the co-generation of electricity, process steam or district heating. The modula concepR rHT characterizes ti face th t y dthab t standardized reactor units with power ratings of 200 MJ/s (so- called modules) can be combined to form power plants with a higher power rating. Consequently the special safety features of small high-temperature reactors (HTR alse ar )o availablt ea higher power plant ratings. Due to its universal applicability and excellent safety features modulae ,th poweR rHT r plan suitabls ti erectior efo n no any site, but particularly on sites near other industrial plants or in densely populated areas. The principle safety feature of the HTR-Module is based on the fact that, even in the case of failure of all active cooling systems and complete loss of coolant, the fuel element temperatures remain within limits at which there is practically no release of radioactive fission products from the fuel elements. selectee th o t de reactoThidu s si r design which guarantees that the modula poweR rHT r plant doe present sno hazary e tan th o dt environment either during normal operation or in the case of any accidents.

The most important design features are as follows :

sphericaf o e us e l Th fuel elements, whic capable har f eo retaining all radiologically relevant fission products up to fuel element temperature approxf so . 160. 0°C The reactor core is designed such that a maximum fuel element temperatur exceedet no 160f s eo i 0dC ° duriny gan accident. Active core coolinnecessart no s gi decar yfo y hea- tre moval during accidents. It is quite sufficient to discharge the decay heat by means of passive heat transport mechanism (such as heat conduction, radiation, natural convection simpla o )t e surface cooler with water flowing

267 through it. This is installed outside the reactor pressure vessel in the primary cell. Reactor shutdown solel absorbey yb r- re element e th n si flector, which, on demand, can drop freely into boreholes. Sole use of Graphite in core areas with high temperatures (fuel elements, core internals). Temperature-incurred failure of this material is impossible at the maximum occurring temperature of 1600 °C. The single phase noble gas helium, which is neutral from a chemical and neutron physical viewpoint, is used as coolant. hige th h o activitt e Du y retentio fuee th l n ni elements , a pressure-tight reactor building is not necessary. The reactor buildin accessibls gi repair efo ry woran t ka time after accidentactivitw lo resule a th s ys a f to release. Reactor corstead ean m generato installee rar sepan i d - rate steel pressure vessels in such a way that there is no danger of component overheating in the case of failure primare ofth y circuit cooling. This installation also increases the accessibility of the component maintenancr sfo repaird ean .

2 NUCLEAR STEAM GENERATING SYSTEM Constructio Desigd nan n Features The nuclear steam generating system (HTR-Module) basically consists of:

The reactor pressure vessel with core, core internals, shutdown systems and facilities for the charging and dis- charging of fuel elements The connecting pressure vessel with hot gas duct The steam generator with the heating tube bundle and the primary circuit blower.

Each HTR-Module is installed in a primary cell, the concrete wall whicf so h suppor reactoweighe e tth th f to r pressure vessel, the connecting pressure vessel and the steam generator pressure vesse theid lan r internals (see Fig. .1) Pipes with water flowing through them (surface coolers) are in- stalleinside th n primaredo e walth f lo y cell neareactoe th r r pressure vessel. These surface coolers discharge dissipated heat during normal operatio decad nan y heat during reactor shutdown.

As shown in Fig. 2 the reactor and the steam generator are positioned beside each other.

This characteristic HTR-Module layout offers substantia- lad vantages :

Afte reactora r shutdown harmfuo thern s ei l natural circulation of hot helium through the primary circuit thanks to the thermohydraulic decoupling of heat source and heat sink. Consequently neeo ther n cooo steae t ds ei lth m generator after shutdown. Thus it can be shut off and left in a hot condition.

268 1 Pebble bed Pressur2 e vessel 3 Fuel element discharge 4 Small sphere shutdown unit Reflecto5 d rro Fue6 l element loading 7 Steam generator tube bundle Stea8 m generator shroud 9 Feed water line 10 Live steam line 11 Primary circuit blower 12 Hot gas duct 13 Surface cooler

^ . . * $ L—y-p « asss^ilüQa — _ — rSg j

1 Reactor pressure vessel Stea2 m generator pressure vessel Connectin3 g pressure vessel 4 Primary circuit blower Primar5 y cell 6 Protective shell n 7 Surface cooler

to OS . Vertica1 FIG. l cross sectio reactoe th f no r building FI 2 GPrimar y circui HTR-Moduln a f to e The positioning of the steam generator beside and lower than the reactor permits simple and operationally favorable upward evaporation. separatioe Th reactoe th f steae no r th cormd generatoean r by the shielding concrete walls of the reactor cell makes it possible to repair all defects in the steam generator bundle or the primary circuit blower without accessibility problems. The staggered arrangement leadfavorabla o st e height-to- width reacto e ratith f o r building. This simplifies verificatio stabilits it case f loadf th no eo n yi s resulting from external impacts (e.g. earthquake, explosion blast wave).

FueThR eHT l Element

Fissionable uraniu presens mi HTR-Module th n ti spherican ei l fuel elements. Each fuel element has a diameter of six centi- metre containd san s approximately 11600 coated particles within the inner graphite matrix. Each of these coated particles consists of a fuel kernel (uranium dioxide UCU) with a diameter of about 0.5 mm, which is coated with several layer pyrocarbof so n (PyCsilicod an ) n carbide (SiC). These layers enclos fuee eth l kernel, thus preventing a fission product release from the fuel kernel.

One fuel element contain uraniuf o g stotaa 7 m f witlo ha U 235 enrichment of 7.8 %. During normal operation the fuel elements attain a maximum temperatur approximatelf eo . °C 0 y85 The evaluatio numerouf no s heat-up experiment shows sha n that the coatings of the particles are capable of practically completely retainin rediologicalll gal y relevant fission products in the intact coated particles up to fuel element temperatures of around 1600 °C. For this reason, the HTR-Module was designed in such a way that the maximum fuel element temperatures limit themselves (i.e. without the intervention of active systems) to values of approximately 1600 °C during all accidents. At temperatures of less than 1600 °C there is only a slight activity release from fuel particles with defective coatings. Numerous experiments have shown, however, thanumbee tth f ro defective coated particle expectee b o st extremels i d y low. This fact also explains the low coolant activity and the low environmental exposur case losf th eo coolanf n so ei t accidents.

Reactor Core

The reactor core, whic locates hi d insid reactoe eth r pressure vessel, consist approx0 sphericaf so 00 0 .36 l fuel elementn si a loose pebble bed with a diameter of approx. 3 m and an average height of approx. 9.4 m. It is cooled by^helium. The mean power densitmeacore e th th limites ne i f d MW/3 yo cor an o mdet outlet temperatur. °C 0 70 o et

270 reactodesige e th Th f no r corfollowine bases th ei n do g principles:

In all accidents and accident combinations, a maximum fuel element temperature of approx. 1600 °C is not exceeded even without active remova decae th yf l o hea t frocoree th m . Decay heat removaeffectee b n lca d solel heay yb t conduction, heat radiation and natural convection to the surface coolers positioned outside the reactor pressure vessel. The reactoshue b tn rdowca n purel droppiny - yb ab e gth sorbers into the reflector boreholes. The uranium loading of the fuel elements amounting to 7 g of uraniu selectes mi d such that water ingress inte oth primary circui resul e accidenn a th f s to a t will cause a lower reactivity increase than the accidental withdrawal of all reflector rods. Therefore the faulty withdrawal of all reflector rods is a covering reactivity accident. This accident is controlled by simply switchin primare th f ygof circuit blower, wherby the permissible fuel element temperatur approxf eo . 1600C ° is not exceeded.

In orde obtaio t r mose nth t uniform possible power density distribution sphericae ,th l fuel elements pass throug core hth e approximately 15 times before reaching their final burn-up.

The ceramic core structure, which essentially consists of side, bottom and top reflectors, forms the cylindrical vessel which holds the pebble bed. The ring-shaped side reflector consists of 24 individual columns made up of single graphite and carbon blocks. The top reflector consists of several layers of graphite and carbon segments cole plenus .p Th dga to locates e mi th n di reflector bottoe .Th m reflector, which also consist severaf so l layers, contains the hot gas plenum. A metallic core vessel, which is supported in the reactor pressure vessel at a point above the connecting pressure vessel, acts as supporting structur ceramie th r cefo core internals. core covee th e Th f vessero constructes li radiatioa s da n shield (so-called top thermal shield) to provide access to the area abov core eth e vesse maintenancr lfo e work. This are filles ai d with stagnant helium during operation.

Th eshutdowe driveth r controd sfo nan l system mountee sar d on the thermal shield. These consists of the reflector rods and the small sphere shutdown units. The 6 reflector rods are for the reactor control and hot shutdown of the core. Each rod consists of several individual sections holdin absorbee gth r materia. C lB In order to initiate a reactor scram, the power supply to the drive is interrupted, thus causing the rod to drop freely into its lowest position in the boreholes of the side reflector due to gravity. 18 small sphere shutdown units serv compensato et increase eth e in reactivity when running dowreactoe nth cola o drt condition. Graphite spheres witcontenC approx n hdiameteB^ a a % d t0 an .1 r o fusee approxshutdows ar d a m m 0 .1 n elements spherese .Th , which are stored in storage containers located above the top

271 thermal shield and over the side reflector, drop freely into the reflector boreholes on demand. During normal operatio primare coolepebbls e i th nth d y db ebe y coolant flow. After flowing upwards through boreholee th n si side reflector, the primary coolant is collected and deflected reflectorp to e ith n thet .I n flows downwards througp to e hth reflector corebottoe e th th ,f d mo pebble whereb an ,th d ebe y it remove heae sth t generate coree th ,n i dbefor e being collect- plenus ga conveyed t man e steaho th e mo th dt generato n ei d a rvi penetratioa side th e n nreflectori .

Steam Generation

The steam generato designes ri once-througha s da , helical tube steam generator. intd heliut steae fe ho o th s e mi m Th generator abov heate eth - ing tube bundle. While flowing aroun steae dth m generator heating tubes, the helium releases its heat to the water/steam side, whereby it cools down from approx. 700 °C to approx. 250 °C. On leavin heatine gth g tube bundle floe ,deflecteth s w i d through 180 °C. The cold helium then flows upwards between the steam generator shroud and the inner wall of the pressure vessel tha o pressurs e tth e vesse cole cooles li th d y dgasb .

The primary circuit blower returns the helium to the reactor throug annulae hth betweep rga connectine nth g pressure vessel and the hot gas duct. The feedwater enters the feedwater nozzle with a temperature of 170 °C. The water flows through the helical tubes from the bot- tope th ,o durint m to g which proces evaporatet si subs i d - san sequently superheated compensatina a .Vi g bundle abov heate th e - ing tube bundle, the steam which is 530 °C hot enters the live- steam nozzl passed ean s throug live-steae hth me th lin o et machine hall.

Primary Circuit Blower

primare Th y circuit blower consist single-staga f so e radial compressor. A speed-controlled asynchronous motor is provided as drive motoe . Th coole s ri d with water. The blower with motor is installed in the blower pressure vessel section as a slide-in unit with vertical shaft and hanging impeller. A blower flap is located on the suction side of the blower. blowee Th r increase pressure heliusth e th f emo sucke fron di m the steam generator outlet side by approx. 1.5 bar.

Cavity Cooler

The cavity cooler surrounds the reactor pressure vessel at a distance of about 1.5 m. It is installed as a closed tube wall in front of the concrete walls of the reactor cell with a clearance distance of approximately 10 cm. Water at a temperature of flowabouC ° 0 st4 throug surface hth e pressurecoolew lo t ra . Its remov o tast dissipates e ki eth d heat amountin approxo gt . connectee th o t W dk intermediat0 40 e cooling systems during normal operation and to remove the decay heat from the reactor

272 of up to approx. 850 kW to these systems in the case of failure maie ofth n heat sink.

Nuclear Ventilation System

reactoe Inth r buildin reactod gan r auxiliary buildine th f go modula poweR rHT r plant areal accessibl,e al sar e during normal operation with the exception of the primary cells. Treated int d reactoe fe ooutsid th s i rr buildineai removo gt e eth dissipate retaido t hea necessare d nth tan y qualitaire th .f yo desiree Th d subpressur maintaines ei regulatiny db supple gth y air flow while keeping the exhaust air quantity constant. During normal operatio releases i environmene ven e r th nth t ai o dt t activitw unfilterelo vene e th th yt o a pollutionstact dvi e kdu .

Dust-bound radioactive material entey vens ma e durin r rth tai g repai maintenancr ro primaree worth n ko y circui connecter to d systems. For this reason, the vent air filter system (aerosol filter) is switched on during such work as a precautionary measure. resule th leakages f ta o , If componentn si s containing primary coolant, gas-borne activity manages to enter the reactor building or the reactor auxiliary building, the facility is automatically switched over to the subpressure maintenance system with its activated corbon and aerosol filters.

3 MODULASAFETE TH POWER F YO RHT R PLANT

3.1 GENERAL

The overriding aim of the safety design is the in-plant reten- radioactive tioth f no e materials which unavoidablyo t fore mdu nuclear fission such that any danger to the environment can be excluded both during normal operation and after accidents. The modula poweR rHT r plan designes ti thay sucn l wa di tal ha accidents based on physically and technically plausible assump- leat tioninadmissiblo no dt o sd e radioactivity releases. This is optimu e primarilth o t m e exploitatioydu favorable th f no e inherent feature smalf so l gas/graphite reactors. Inherent Safety

"Inherent (i.e. intrinsic) safety" is a specialized term used in this case to describe the fact that the reactor itself reacts to certain malfunctions withou actuatioe tth activf no e systems or external controlling interventions in such a way that no inadmissibl ever eo n dangerous situation reachede b n sca . These reactions are governed by the laws of nature, i.e. they always function regardles conditioe th f so activf no e systems. This means that they cannot malfunction or fail. The technica nuclead lan r physical desigHTR-Module th f no s ei such that the fuel element temperature always stabilizes itself approxt a . 160case eveth assumef C 0 eo ° n ni d failurl al f eo active shutdown and decay heat removal systems.

273 hane on d e thiO th nachieves si face th t HTR e y thatdb th -n ,i Module, thertemperatura s ei e spa approxf no betweeK 0 e .70 nth maximum permissibl fuee th l f eelemeno t temperatur maxie th -d ean mum operating temperature of the fuel elements. This temperature span ensures tha reactoe tth r core shuts itsele f th dow a nvi negative temperature coefficients of reactivity, even after accident-incurred introduction of the existing surplus reactivity, before the above limit temperature of 1600 °C is reached.

On the other hand, the selection of a low mean power density in the reactor core, the selection of a suitable geometry for the reactor core and the surrounding core internals, and the use of appropriate materials ensure that the decay heat can escape froreactoe mth rsurroundine corth o et g component strucd san - tures solely by means of physical processes (heat conduction, radiation, convection).

Activity Release Barriers

A characteristic safety featur HTR-Module th foune b f eo o dt s ei in the fact that, in all operating and accident situations, the radioactive materials formed during nuclear fission are enclosed in the coated particles in such a way that a significant activity release from the coated particles can be excluded.

This safe activity enclosur guarantees ei desige th y dnb of the fuel particle coatings and the inherent limitation of the maximum possible accident temperatur fuee th l f eelemento s to approx. 1600 °C. The radioactive materials escaping from the few defective particles are partly retained in the fuel element matrix. The non- retained fraction is transferred to the primary coolant and is distribute primare th n di y circuit. The gas-borne activity in the primary circuit is reduced by radioactive decay separatioy ,b heliue th n mni purification facility and by deposition on the surfaces of the primary circuit. The primary circuit therefore acts as a further barrier to preven releasta radioactivf eo e materials. Leakagee th n si connecting pipes which canno isolatee tb vere dar y improbable plannee th duo et d quality assurance measures. Inonthelesa n s postulated leakage principln ,i e onlvere th yy low gas-borne primary coolant activity and a small fraction of activite th y deposite primare th n do y circuit surface coule db released int reactoe oth r building thir .Fo s reason basicd ,an - ally becaus hige th h f eretentioo n capabilit coatee th f dyo particles tightneso ,n s requirement imposee reactoe sar th n do r building of the HTR-Module for adherence to the permissible accident dose limit value Germae th Articlf s o f no Radiatio8 e2 n Protection Ordinance, i.e releasee .th d activit dise b n -yca chargeenvironmene th o dt t withou dangery tan .

3.2 ACCIDENTS

Measures for Accident Control

Even thoug nucleae hth r power plan designes ti constructed dan d in accordance with the highest quality requirements, it is im- possibl completelo et y exclude failur technicaf eo l systems.

274 selectee th Duo et d HTR-Moduledesige th f no ,prow onlfe -ya tective action stakee b hav o nt e after accident shuo st e tth reacto safa n erkeei o t conditiondowt pi d nan . After being shut down HTR-Module ,th lefe b tn estandinca a n i g conditiot ho thin i s, nconditionas maximue ,th m fuel element temperatures only increase to approx. 1100 °C at a normal operating pressure even without active core cooling, and to approx. 1600 °C after loss of pressure.

e steaTh m generato thermohydraue th o coolee rt b neee t ddu no d - lic decoupling of reactor and steam generator after shutdown. The HTR-Modul lefe b tn conditiot estandinca ho a n gi n until the causes and consequences of the accident have been repaired. As there is no need to remove the decay heat by means of forced circulation withi primare nth y circuit separato ,n e decay heat removal ciruit necessarye sar . In the case of an accident, the decay heat is removed from the HTR-Module via heat conduction and heat radiation to a cavity cooler installed outside the reactor pressure vessel. This cooler is capable of protecting the core internals, the shutdown systems, the reactor pressure vessel and the concrete structure of the reactor cell from inadmissible temperatures during all accidents. During normal operation cavite ,th y cooler remove dissipate sth - ed heat from the reactor cell. It is therefore continuously in operatio specialle b need t nan dno y starte whe p accidende u nth t occurs. The water through-put and water temperature of the cooler are so low that, for an assumed failure of all pumps or recooling chains, water supplie hosa y eb d pipe after approx hour5 .1 s (e.gy .b the fire department) would suffic ensuro et e adequate cooling of the reactor cell and the reactor pressure vessel without any damage having been plane causeth tn di tha t woul allot dno w further operation. In principle postulatecasa e f th e o n ,i d primary circuit leakage, it is possible to directly discharge the primary coolant unfiltered to the environment. Nevertheless, in the case of depres-_ surization accidents with leakage, cross-sectionm c 2 o t p u f so it is planned to pass the leaking primary coolant through the subpressure maintenance system and to discharge it filtered via the stack in order to minimize the radiological exposure of the environment. In the case of larger assumed leakage cross-sections which would caus notabla e e pressure increas reactoe th n ei r building buildine ,th shortls gi y depressurized with direct dis- environmente chargth o t e .

Reactor Protective Actions

Accidents cause characteristic changeprocese th n si s variables whic detectee reactoe har th y b dr protection syste whicd man h initiates shutdown of the modular HTR plant if the specified limit values are exceeded. Regardless of the accident, the same three protective actions are always performe planr dfo t shutdown:

Dropping of the reflector rods - Shutdown of the primary circuit blower Isolation of the steam generator.

275 The firs nucleaactionr o tfo tw e rsar shutdown purposes, while isolation of the steam generator (i.e. closure of the feedwater and live-steam valves) separates the reactor plant from the steam power plant. In the case of a loss of pressure accident and a steam generator heating tube rupture primare ,th y circuit isolation valvee sar additionally closed resp steae .th m generato quickls ri y drained.

3.3 BASIC DESIGN ACCIDENTS

In orde asseso rt efficience sth safete th f yyo e desigth d nan planned protective measures occurrence ,th accidentf eo s si assumed in analyses and the resulting accident courses are examined.

Reactivity Accidents

In orde defino rt coverinea g reactivity acciden postus i t ti - lated shutdow 6 thal tal n rods withdraw with maximum speed during full load operation due to an assumed fault in the control system.

The reactor protection system detecte accidene th sth o t e tdu neutrochangtempere s th ga n t ei n- ho increase fluth e th r x o n ei ature and shuts the reactor down once it reaches the corresponding limit values. This doe resulnotablt y sno an n ti e fuel element temperature increase in the core. No activit releases yi reactoe th o dt r buildin environe th r go - ment. All other events, in which a reactivity increase is possible, e.g. faulty operation of a small sphere shutdown unit, water ingress, erroneous raising of the primary coolant throughput or erroneous lowerin cole s temperatureth dga f go coveree ,ar d by this accident.

Disturbances in the Primary Heat Transfer System

This covers all events leading to a disturbance of the heat transport form the reactor core via the primary circuit and the water/steam circui condensee th procese o th t r ro s steam converter . Such eventcausee b n sdca interruption e.ga y .b floe th w f no throug primare hth y circuit (e.g. failurprimare th f eo y circuit blower througr )o secondare hth y circuit (e.g. pump failure) or by possible changes in the heat balance (e.g. faults in the live-stea procesr mo s steam extraction). Los powef so r (emergency power case) also leaddisturbanca o st e of the heat transfer, as the primary circuit blower and the feedwater pumps simultaneously lose their driving energy.

mase Th s primare flowth secondard f so an y y circuit monie sar - tored by the reactor protection system. If the values exceed or fall short of a given flow ratio the plant is automatically shut down. Sudden changes in the heat removed are also made eviden increasn a cole temperatures y th tb dga n ei . Oncea specified limit value is reached, the plant is also shut down reactovie ath r protection system.

276 The sole consequence of all these disturbances and accidents is the temperature loading of components, but these loadings are lower than the permissible limit values. The shutdown reactor remains in a hot condition until the disturbance is eliminated. The maximum fuel element temperatures are around 1100 °C. Non thesf eo e events result activitn a n si y releasreace th -o et tor building or the evironment.

Leakages and Ruptures on the Secondary Side

These events includ leakagel eal ruptured san e tuben si th f so water/steam circuit. The rupture of a live-steam line in the reactor building is the covering accident. The rapidly sinking live-stea rupturme pressurth o t e e causeedu s a fast increase in the feedwater transport and, as a result, the flow ratio between primary and secondary circuit is disturbed. The reactor protection system initiate protective sth e actions droppin reflectoe th f go r rods, shutdow primare th f no y circuit blower and isolation of the steam generator once the limit value has been reached. In the case of a leakage which cannot be isolated, the steam generator slowly steams out. outflowine Th g steam escapes frobuildin e pressure mth th a gvi e relief openings. inadmissiblo Thern e ear e pressur temperaturr eo e loadingf so the reactor building or its internals. steae Th m generato internals it d designeran e sar withstano dt d the loadings resulting from this event. Consequently a heating tube rupture is not to be assumed as a subsequent failure. Leakages and ruptures on the secondary side therefore only lead to temperatures and pressure loadings of the components and the building. There is no activity release to the reactor building or the environment.

Steam Generator Leakages (Water Ingress)

In the assumed rupture of a steam generator heating tube water or steam forces its way into the primary circuit due to the overpressur secondare th n eo y sidbar)0 6 ee o t (19Th . r 0ba water ingress is detected by means of a moisture measurement.

reactoe Ith f r protection limit valu exceededs ei reactoe ,th r is shut down and, in addition, the steam generator is rapidly draine limio dt quantite tth watef yo r enterin primare gth y circuit. The water whic leakes hha d intprimare oth y circui removes ti d fro usint mi heliue gth m purification facility wit watee hth r separator. Thi switches si manuallyn do . Water entering the primary circuit can cause corrosion of the fuel elements due to the reaction HO + C ^ H + CO. Of the approx. 600 kg of water, which at most enter the primary circuit during this accident sequence, approxconverg k 0 .10 t coursconter-measuree e tth th o n f wateei o s rga s described above. The remaining quantity of water and the forming water gas are removed from the primary graphite burnup at the fuel elements as a result of the water gas reaction amounts to approx. 0.7 %. This is non-problematic both with regard to fuel element corrosion, stabilit fissiod yan n product release.

277 If failurwatee th rf e o separato postulateds ri primare ,th y circuit pressure control alon alss ei o capabl preventinf eo g an increas pressurn ei e resulting froadditionae mth l formation of water gas. Consequently the pressure relief system of the primary circuit is not actuated in this case either.

The sequences described above which take place due to the ingress of water do not result in any activity release from the primary circui reactoe th o t r buildin environmente th r go .

If the improbable accident combination: "Failure of the water separator and failure of pressure control" is assumed, a pressur primare th ris n ei y circui occurn tca . This acutatee sth pressure relief system after abou hourst5 . Aresultsa primar,e th som f eo y coolant mixed with stead man water gas is blown off into the reactor building. It is then filtered and released to the environment from this building.

The radiological consequence coveree consequencee sar th y db s described for the accident "Loss of pressure in the primary circuit". They remain under the values stipulated in Article 28 of the German Radiation Protection Ordinance by approximately four orders of magnitude.

Loss of Pressure in the Primary Circuit With Subsequent Core Heat-Up

Possible events initiating this acciden primariln tca leake yb - ages in a system connected to the pressure vessel, as the result of which primary coolant can flow into the reactor building.

Primary coolant leakages are detected by the reactor protection system by way of the decreasing primary circuit pressure. The reactor protective actions droppin reflectoe th f go r rods, shut- primare th dow f no y circuit blower, closurprimare th f eo y circuit isolation valves are initiated as counter-measures. Closur thesf eo e primary circuit isolation valves, whice har installe connectine th n di g pressur e lineth o st e vessel outside the primary cells, prevents a further discharge of primary coolant. If helium resp. activity is measured in the vent air, the plant is automatically switched over to the subpressure maintenance system to ensure that small leakages are released in filtered form. This measure is taken to minimize the environmental exposure . In orde determino rt maximue eth m consequences ,totaa l non-iso- lable rupturlargese th f eo t connectin directl) m gc lin3 y(3 e pressure atth e vesse assumes li coverins da g accident.

Rupture of the line is followed by the complete depressurization of the primary circuit until it has reached the same pressure as the environment after several minutes. In order to prevent an inadmissibly high internal pressure in the reactor building, the primary coolant is discharged to the environment via pressure relief openings with flaps in the reactor building.

The environmental exposure radioactive causeth y db e materials released with the escaping primary coolant is much lower than

278 the permissible accident limit values given in the German Radia- tion Protection Ordinance. After depressurization core ,th e slowly e beginth f heai o s t p tu decay heat cannot be removed by the main heat sink. After approx. 32 hours maximu,a m fuel element temperatur approxf eo . 1600C ° is reached in one small sector of the reactor core, as shown in Fig. 3. Less than 2 % of all fuel elements are exposed to a temperature exceeding 1500 °C. The temperatures of the other fuel elements are much lower (see Fig. 4) .

side reflector absorber position

0 20 40 60 80 100 120 140 160 180 200 1 h FIG ; Temperatur3 . e variations after depressurization accident with core heat up

0 20 0 18 0 16 0 14 0 12 0 10 0 8 0 6 0 4 0 2 0 FIG: TimedependenA- . t fractio f fueno l elemen variour tfo s temperatures (depressurization with core heat up)

279 relatee th core d th heatine dean th f expansio o o t p gu e Du f no the cooling gas, radioactive materials can be released into the reactor building from the primary circuit. However, as the limit temperature for fuel elements of approx. 1600 °C is not exceeded, there is no massive fission product release from the core. If activit measures yi vene th t n di subpressuraire th , e maintenance system with its filters isautomatically started. Consequently the radiological consequences of the activity releas keplevea e ear t t a l whic mucs hi h lower thalimie nth t values given in the german Radiation Protection Ordinance (see Table 1). In addition face ,th t tha core tth e only heat slowlp su y provides adequat implementatioe eth timr efo counter-measuresf no . reactoe Th r buildin alwayn gca enteree sb staro t d t repair measures both durin afted gan r core heat-up.

Table 1 shows the maximum radiation exposure of single organs in compariso limie th to nt value s listee Articlth n di f o 8 e2 German Radiation Protection Ordinance. The calculation rules of this article include alle exposure paths like submersion, inhalation and ingestion (ingestion over a period of 50 years !).

Tabl: 1 e Accident doses afte losra pressurf so e accident with core heat-up

Organ Accident doses Limit values r em after Art8 2 . RPO rem

Bone 0.0081 30

Liver 0.012 15

Total body 0.012 5 Thyroid 0.47 15

Kidneys 0.012 15

Lungs 0.012 15

Gastro-intestinal 0.018 15 tract

Skin 0.0034 30

The table shows that, even in this covering accident with activity release valuee ,th muce sar h lower thalimie nth t values of the German Radiation Protection Ordinance and that there is no danger to the population.

280 3.4 PROTECTION AGAINST EXTERNAL IMPACTS

In additio acciden, nto t events, whic causee har operatioy db n reactoe oth f r itself, other event observede sar . These havn ea externa lcausn planimpacca e eth d tdamagean n t o .

Earthquake Important safety-relevant systems of the modular HTR power plant are designe maximua r dfo m seismic intensity. Accordino gt scientific knowledge it is extremely improbable that this intensity will be exceeded. The maximum possible loading as the result of an earthquake is taken into consideration in the design of the reactor building and the central control building. Insid reactoe eth r building basically onlpressure yth e vessel with primary and secondary circuit isolation valves, the cavity coolers with safeguarded recoolin systeml al d sgan required for correct functioning of the reactor protection system are designed to withstand an earthquake. The systems installed in the central control building such as the reactor protection system emergence ,th y power system controe ,th l room etc. also function after an earthquake due to suitable dimensioning. This means that all measures required for minimization of the activity relase can be taken.

Aircraft Crash

The reactor buildin constuctes gi d such tha crashinta g aircraft cannot penetrate the roof and the outer walls. The pressure vessel including the primary circuit isolation cavite valveth d ys an cooler designee sar withstano dt e dth vibrations transferred to the building as the result of a crash and remain intact. If defects arise in the systems connected to the primary circuit after the aircraft crash, the reactors are automatically shut down and isolated by the reactor protection system as the result disturbancee ofth decae th yd ,heaan cavitremovee s ti th a ydvi coolers. If recoolin thesf go e coolers wer failo et , water coul suppe db - lied to the surface coolers via fire department hose pipes. More tha hour0 n1 s tim availabls ei initiatioe th r efo thif no s measure. centrae Ifth l control buildin destroyes gi aircrafe th y db t crash necessare ,th y reactor protective actions will nevertheless bdesige th actuate o nt e (e.gddu . closed circuit current principle) and the reactor will be brought to a safe condition. The plan subsequentln tca monitoree yb d fro emergencma y control room located in side the reactor building.

Explosion Blast Wave

The reactor building together wit safety-relevane hth t systems suc pressurs ha e vessel with primary circuit isolation valves and cavity cooler designes si withstano dt loadinge dth - sre sulting from an explosion blast wave. If other parts of the plant are damaged the reactor is automatically shu reactote dowth y nb r protection system.

281 If the central control building is inoperative as the result of the explosion blast wave, the plant can be monitored from the emergency control room located inside the reactor building. In the case of failure of the recooling of the cavity coolers, suppliee wateb surface n th rca o dt e cooler fira svi e department hose pipes. More than 10 hours time is available for the initiation of this measures.

4 MAIN DATA

Overall plant (2 reactors)

Reactor termal power 400 ""th Generator terminal power - for max. process steam 72 MW generation - for min. process steam 124 MW generation

Quantity of process steam max. 115 kg/s min. 47 kg/s

Process steam pressure/ 17 bar/27C 2° temperature

Reactor

Numbe fuef ro l elements 360 000

Fuel element diameter 6 cm

Fuel element charging process Multiple passage

Primary coolant Helium

Helium flow at 100 % power 85 kg/s

Helium temperatures 250/70C ° 0

Mean helium pressure r ba 0 6

Mean power density 3.0 MW/nf Steam Generator

Live-steam pressure at SG outlet 190 bar

Live-steam temperature at 530 °C SG outlet

Live-steam mass flow 77 kg/s

Feedwater inlet temperature 170 °C

282 MHTGR LICENSING APPROACH AND PLANT RESPONSE TO OFF NORMAL EVENTS

A.J. NEYLAN, F.A. SILADY GA Technologies, Inc., San Diego, California, United States of America

Abstract

The MHTGR design meets stringent top-level regulator d usean yr safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, generic top-level regulatory criteria have been specified from US NEC and EPA sources to guid e designth ee uses furtheTh ha .r r specifiee dth that a t t thesme e b e plant boundary A probabilisti. c risk assessmen s beeha tn utilize o select d t licensing basis events that cove wida r e spectru f eventsmo e eventTh . s have been grouped into three frequency regimes governee 10CFR5th y b d0, I App. 10CFR100 G ris dosd PA k an d e ,an criteria .

The MHTGR fues beeha l n designe o limi t de primar th t y circuit activitieo t s levels that, even if completely released, are within those allowed by 10CFR100. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attacky controllinb d an , g heat generation e corTh e. geometry, core power, core power density, heat removal geometry, and heat sinks are designed to provide passive core temperatures during accidents. To limit the potential for chemical attack of the MHTGR core, the primary coolant boundary is designed to make large ingress of air and water very unlikely. Furthermore, the fuel particle coating e highlar s y imperviou o oxidizint s g agents tha o ented te th r primary coolant. The core heat generation is controlled by the large negative temperature coefficient and by insertion of control material to maintain a subcritical core configuration during the span of the accidents. Thus, the thre y safetke e y functions have been accomplishe y relyine inherenb d th n o g t characteristic e MHTG th o tha s e Rresponsf o th st s largeli e y passivd an e relatively simple.

SAFETY PHILOSOPHY

The overall safety philosophy guiding the design of the MHTGR is to produce a safe, economical plant design whic d huse an rmeet C requirementNR s y b s providing defense-in-depth throug e pursuith h f fouo t r goals) Maintai1 : n Plant Operation, 2) Maintain Plant Protection, 3) Maintain Control of Radionuclide Release Maintai) 4 d ,an n Emergency Preparedness.

With e achievemenregarth o C criterit e accomplishmendNR th f r o t fo a e th f o t first two goals, measures are taken in the design of the MHTGR to minimize defects in the fuel so that normal operational releases or any accidental release f primaro s y circui d workean t w ractivit lo exposuree ar ye ar s minimized.

283 The unique aspect of the MHTGR, however, is the approach which has been taken to achiev thire eth d goa therebd lan y minimiz desige eth n requirements froe mth fourth goal. To accomplish this with high assurance, the design of the MHTGR has been guided by the additional philosophy that control of radionuclide releases be accomplished by retention of radionuclides within the fuel particles with minimal reliance on active design features or operator actions. e overalTh l inteno providt s i t a simple e safety case that will provide high confidence that the safety criteria are met. This approach is consistent with the NRC's Policy on Advanced Reactors (Ref. l). There are two key elements to this philosophy which have had a profound impact on the design of the MHTGR (especially in the selection of core size and geometry, power density, and vessel type); the basis for each element is described below.

First, the philosophy requires that control of radionuclides be accomplished with minimal reliance on active systems or operator actions. By minimizing the, need to rely.on active systems or operator actions, the safety case centers on e e behaviointegrit e th th law th f physicn o s o f f o passivrd o y an s e design features. Arguments need not center on an assessment of the reliability of pumps, valves and their associated services or on the probability of an operator taking various actions, give associatee nth d uncertainties involven i d such assessments.

Second, the philosophy requires control of releases by the retention of radionuclides within the coated fuel particle rather than reliance on secondary barriers (such as the primary coolant boundary or the reactor building). Proof f containmeno s dramaticalli t y simplifie f argumenti d n centeca s n issueo r s associated with fuel particle coating integrity alone.

TOP-LEVEL REGULATORY CRITERI USED AAN R SAFETY REQUIREMENTS

Top-level criteria and requirements are defined primarily from two sources: e regulatorth , whose concer s primarili n y public healt e d safetyth an h d an , user, whose concern is all encompassing (e.g., safety, performance, availability d economics),an . e fouth Eacr f ho goal s beeha s n quantifiea y b d serie f top-leveo s l criteri d requirementan a e top-leveTh s (Ref. 3) l ., 2 regulatory criteria are the basis for plant licensability with the Preliminary Safety Information Document (PSID) and the Probabilistic Risk Assessment (PRA) presentin desige th w n ho gmeet s these criteria.

284 e followinTh g bases were e selectioadopteth r fo f dtop-leveo n l regulatory criteria.

1. Top-level regulatory criteria should be a necessary and sufficient f direco t tse statement acceptablf so e healt safetd an h y consequences or risk individualo st publice th r so .

2. Top-level regulatory criteria should be independent of reactor type and site.

3. Top-level regulatory criteria should be quantifiable. e firsTh t basis ensures thae criterith t e fundamentaaar protectioe th o lt f no the public and the environment. The second, consistent with the first, requires that the criteria be stated in terms which do not discriminate among reactor types and sites. Finally, the third basis ensures that compliance with the selected criteri demonstratee b n aca d through measuremen calculationr to .

Through comparison wite selectioth h n bases e followinth , g regulatory sources have been foun o contait d n numerically-expressed criteri r limito a s which appropriately form top-level regulatory criteria.

R 2804F 1 4- Polic5 1 . y Statemen n Safeto t Operatioye th Goal r fo sf o n Nuclear Power Plants

2. 10CFR20 - Standards for Protection Against Radiation

3. 10CFR50, Appendi Numerica- I x l Guide Desigr fo s n Objectiveo t . .. s s ReasonablA Meee Criteriw th tLo s y"A a Achievable Radioactivr fo " e Materia Effluentn i . .. l s

4. 40CFR190 - Environmental Radiation Protection Standards for Nuclear Power Operations

5. 10CFR100 - Reactor Site Criteria

. 6 EPA-520/1-75-00 - Manua1 f Protectivlo e Action Guide r Protectivfo s e Actions for Nuclear Incidents

The numerical consequence or risk values contained within the above regulatory top-levee sourceth e ar s l regulatory criteriMHTGe th Rr fo a(Ref . l) .

285 utility/usee Th r grou s specifieha p n additionaa d l safety requirement (Ref) 2 . thas mori t e restrictiv n thai e t itee top-leve th 6 abovm f o e l regulatory criteria is to be satisfied at the plant boundary. In this way the emergency planning zone, whic'h is generally 16000 m ( 10 miles) for US LWRs, is reduced to the MHTGR' m Exclusio 5 42 s n Area Boundary (EAB). This allow utility/usee sth r o limit t emergency e ared personnel..-withidrillth an a o t s s controlit n e Th . need for offsite sheltering and evacuation is obviated, the public's normal day-to-day activitie t disturbe no e proximit e th e MHTG ar sy th b d Rf o yplant . e specifiTh c quantitative user A ProtectivrequirementEP e th e e ar sActio n Guideline wholm re sthyroim e1 re (PAGsbodd 5 an dyf )o dose s evaluatee th t a d 425 m.

LICENSING BASIS EVENTS

For the purpose of deriving the regulatory licensing bases of the MHTGR, the probabilistic bases for the design have been cast in a framework and format simila o tha t rf traditiona o t l licensing approaches. f Postulatioo t se a f no bounding licensing basis events is one of the key elements in the traditional US regulatory process. Licensing basis event e useo ar demonstratt sd e compliance with dose criteria for a spectrum of off-normal events.

10' pr

APPENDIX I 10°

10l-l

0 1 x 5 2. i-Z 10-2 3 a. cc vu a. 101-3 0 10 R '10CF

a tu r« -4 cc 10 10 L*-USER EMERGENCY NO EVACUATION 10-5 PLANNING I REQUIREMENT BASIS REGION NRC ACUTE FATALITY SAFETY GOAL-——""

5x10 -7

...I l i .1 . Mil 10" 10" 10-1 10° 10' 10J MEAN WHOLE BODY GAMM A(REMB DOSEA T |EA

FIG . Licensin1 . g basis regions.

286 e FoMHTGRth r , selectio f licensino n g basis events (LBEse s th basei ) n o d probabilistic risk assessment. The use of the PRA for LBE selection provides a basi r judgingfo s a quantitativ n ,i e manner e frequenc th ,entire th f eo y event sequence and, therefore e appropriatth , e dos r riso e k criteri appliede b o t a .

The initial step in the selection of LBEs is to establish a frequency- consequence risk plot defining three regions bounde n frequenci d y threb y e agreed-upon mean frequencie n consequencei d an s y allocateb , d dose limits relate o 10CFR5t d 0 Appendi , 10CFR100I x e PAGsth .r o , Figur provide1 e s this plot as established for the MHTGR.

Those families of events in the PRA having the potential for radionuclide releases or consequences in excess of those allowed by the top-level regulatory criteria were it not for design selections that function to control the release of radionuclide e thosar s e selecte s LBEsa d . Depending upon their predicted frequency e selecteth , d e followineventth e encompassef ar o s e g on thre y b d e categories :

1. Anticipated Operational Occurrences (AOOs): These are families of events expecte o occut d r plane onc th r mor o etn i elifetime . Their dose consequence e realisticallar s y e analyzePSIo th t D n i d demonstrate compliance with 10CFR50 Appendi. I x

2. Design Basis Events (DBEs): These are families of events lower in frequency than AOOs that are not expected to occur in the lifetime of e plan on t whicbu t h might a occularg n i re populatio f MHTGRo n s (approximately 200). The families of events selected as DBEs at this e MHTGstagth n Ri e desig e listear n n Tabli d . 1 eThes e DBEe ar s evaluated conservativel e PSIth D n i yagains e 10CFR10th t 0 dose criteria.

. 3 Emergency Planning Basis Events (EPBEs): Thes e familiear e f o s events lower in frequency than DBEs that are not expected to occur in the lifetime of a large number of MHTGRs. The EPBE consequences are analyzed emergencr realisticallfo A PR ye planninth n i y g purposed an s environmental protection assessments.

In additio o demonstratint n g compliance wite dosth h e e top-levelimitth f o s l regulatory criteria and the user safety requirements, the LBEs are considered collectively to show compliance with the NRC Policy Statement on Safety Goals . (Ref4) .

287 TABL1 E MHTGR DESIGN BASIS EVENTS

Design Basis Event

Loss of forced core cooling

Main loop transient without contro trid lro p

Control rod withdrawal without main loop cooling

Contro withdrawad lro l without forced core cooling

Earthquake

Moisture inleakage

Moisture inleakage without forced core cooling

Moisture inleakage with moisture monitor failure

Moisture inleakage with steam generator dump failure

Primary coolant leak

Primary coolant leak without forced core cooling

PLANT RESPONSE TO OFF-NORMAL EVENTS

As an introduction to the plant response to off normal events an illustration of the MHTGR safety characteristics is made. To do this, the available fission product inventory is compared to the allowable activity release from the plant that would satisfy the long-term 10CFR100 and PAG thyroid dose limits. To keep the example simple, only the dominant contributor, 1-131, is considered. The allowables for 1-131 release are obtained from the dose limits using a conservative USNRC Regulatory Guid weathe4 1. e r dispersion factor e resulTh . t e 10CFR10G th PA n allowabli r a sd fo an 0 i C e8 1-13d an i 1C releas0 25 f o e thyroid doses of 150 and 5 rem, respectively.

288 In comparison, the MHTGR fission product inventories of 1-131 are given below foequilibriun ra m core:

Maximum Expected Location 1-131 Inventories (Ci)

Core 0 1 x 0 9. Within fuel particles

Vessels Flateout in circuit 20 Circulating 0.018

As show e plateouth n f d abovcirculatino an tl e al eve f i ng activits i y released e totath , l releas n ordea f magnituds o ri e e lower thae 10CFR10th n 0 limits. Therefore, the safety approach is focused logically on limiting the fractional release from fuel particles to less than 3 x 10~ (250 divided by 9 x 106) and to less than 1 x 10"6 {8 divided by 9 x 106) for the 10CFR100 and PAG doses, respectively. e approacTh e coate th ho rele MHTGe desigt n take th th do ys i Rf n fue o ni n l particles for meeting the 10CFR100 doses and on other additional largely passive retention barriers for meeting the more restrictive PAG doses. Three functions have been identified which, when accomplished, assure that radionuclide retention withi fuee nth l remains acceptable:

1. Remove core heat 2. Control chemical attack 3. Control heat generation

Ther mane ar ey ways these function e accomplishedb n ca svarioue th d s,an LBEs utilize different design selections to perform the same function depending upon the accident scenario. Generally, the less frequent LBEs rely more heavily on passive design features. For example, the MHTGR has three independent and diverse coolin gf whic o systems e n certaihi on cand y an ,,an n LBE perfor, sdo m the function of removing core heat. However, while this multiplicity of systems capable of performing these functions contributes to increasing the margi analysesE consideres i MHTGe LB f safet d th no e Ran r th MHTGe fo yn ,th i d R safety design approach emphasize a f minimuslargelo t se my passive design features which, by themselves, are sufficient to accomplish these functions. How the MHTGR meets each of the three key safety functions is now briefly discussed by examining selected LBEs and the fractional release from the fuel.

289 1800

PRESSURIZED

BOO

600 1 1 1 1 19 100 200 300 400 SO TIME PAST REACTOR TRIP |h|

MW(t0 35 ) . FIG2 transient. s during depressurized pressurized an d conduction cooldown.

I.U ...... , .... . y ^

0.8 O

sï o O b o < 0.6 oc u. LU ce o 3 0.4 — S 0

0.2 /NORMAL ,MAX TEMPERATURo E /PEAK FUEL /DURING LICENSING °

/ TEMPERATUR / BASI9 Eo So EVENT o o o S J 1 1 U^ ^„4.^nOQr>oortnP 1 1000 1200 1400 1600 1800 2000 2200 2400 26 TEMPERATURE |°C)

FIG. 3. Coated fuel particles maintain integrity at high temperatures.

290 Remove Core Heat

The inherent feature r heafo st removal includ e intrinsith e c core dimensions and power densitie e reactoth f o sr core, internals passive vessed th ,an d an el cooling pathwa ye environment th froe cor o th mt e . Figur 2 presente e th s temperature transients for two DBEs, one with the primary system pressurized and one depressurized, in which the first two independent means of forced coolin e unavailablear g . Passive heat remova y conductionb l , radiatiod an n natural convection froe corth me throug e reactoe vesseth th h o t lr cavity cooling system limit fuel temperatures to acceptable levels. Passive heat remova e larg th possibls i leo t thermae edu l marginfuele shows th A .n n i sni Figure 3 the fuel must exceed approximately 2100 C before thermal decomposition of the silicon carbide coating results in significant failure. The normal peak fuel temperatur mucs i e h lowe t 110a r . 0C

e Thereforedepressurize th e cas f th o e n i , d conduction cooldown DBEe th , release of 1-131 from the fuel particles is limited by the passive heat removal to 2 x 10 of the fuel inventory. This fraction is further reduced to 0 1 sinc e x releas th e6 s sloi e o thae s woveral th t l release fraction considering only fuel retentio ~ 10CFR1010 x n 3 0meet e allowableth s . Additional passive retention mechanisms within the vessel and building reduce this furtfurthet r to 6 x 10" , thereby also easily meeting the 1 x 10" PAG allowable.

Control Chemical Attack.

e inherenTh t feature r controllinfo s g chemical e fueattac th y wateb l f o kr include the non-reacting cooling, a water-graphite reaction that is endothermic and requires temperatures abov e averagth e e normal operating conditionsd an , the silicon carbide coating e fueth l n o itselfs e MHTGTh .R design features that limit water s consequenceingresit d an s s includ e limiteth e d sourcef o s water, reliable detection and isolation systems, and two forced core cooling systems.

e higTh h e qualitfueth lf o particly e coatings limit e 1-13th s 1 inventory availabl r o watereleast fo e e r du o ethost chemica E eLB ly an attac r fo k particles with initially failed particles (from either in-service failure or manufacturing defects)e ful th e l fractiof Th o . " n10 availablx 2 s i e inventory. Thi s furthei s r reduce a facto y b f e d.01o rth 5l al sinc t no e iodine will be released if the fuel kernel is hydrolyzed. The total retention due to the fuel is then 3 x 10" , meeting the 3 x 10 10CFR100 allowable.

291 Other additional retention mechanisms limit 1-131 release by limiting the amount of water entering the core and by retention in the vessel and -8 surrounding building o givt sn overal a e l release0 1 fractio x 6 f o n which allowableG PA " 10 .x 1 meet e sth

The inherent features for controlling chemical attack of the fuel by air include the non-reacting coolant, the embedded ceramic fuel particles, the nuclear grade vessel, and the below grade vessel silo. Figure 4 presents the fraction of the core graphite reacted for two sizes of primary coolant leaks without forced core coolin e EPBEon g)d an (onconsiderin E eDB e amounth g f o t air available in the reactor cavity silo. As shown the fraction reacted is less than 10- 4 of the core. No oxidation of the embedded fuel particles is predicte y LBEan .r fo d

NT3 o - tUJ l UceJ - gio-Ul 4 f 13 SQ. IN. LEAK ___——-- ——— <

/0.1 LEASQ. IN .K i i i i i i 1 1 1 t \ i i i i i 1 1 1 i i i i i i 1 1 S10-5 J 10° 10" 10 TIME (HRS) PAST

PRIMARY COOLANT LEAK LEADIN CONDUCTIOO GT N COOLDOWN

FIG . Limite4 . d air-graphite reaction retains radionuclide coren si .

Control Heat Generation

The inherent features that control reactivity include a strong negative temperature coefficient .singl,a e phas voio (n ed coefficient neutronicalld )an y inert coolan d largan t e thermal margins. These inherent characteristics cause the reacto o inherentlt r y shutdown s showa A pressurize n .Figr i nfo .5 d conduction cooldown without any control insertion, fuel temperatures remain low. Furthermore, the plant protection system, which is separate from the operational system, include o diverstw s e reactivity control systems thae ar t gravity inserted and highly reliable to protect against even rarer events. The reactivity systems can maintain the core subcritical with margin for the maximum water ingress.

292 lou.u

1200 MAXIMU TRI0 MW/ P 0 S^~^ ————— — - MAXIMU _ _ TRI/ MPW tftul a Ul "°° / —— ———— ce I Se 1000 T AVERAGE W/0 TRIP oc . . . -----. _ - Ul "•*" a. 900 Ul / ^•""' AVERAGE W/ TRIP , >' 000 f.

?nn i i i i i i i i i i i i il 0 2 4 6 8 10 12 14 16 18 20 22 24 26 26 30 TIME HRS

FIG. 5. Acceptable pressurized conduction cooldown temperatures with and without reactor trip.

In conclusion, the passive safety features of the design prevent and mitigate radionuclide release ove wida r e spectru f off-normao m l events. Preliminary analysis has shown that for the release mechanisms and the associated severe accidents considered e regulatorth , y d relyinuseb an yt r me criterig e b n ca a largely on the HTGR fuel performance attributes. Attenuation of release by other barriers such as the reactor vessel and reactor building provides additional margin in meeting the 10CFR100 and PAG dose requirements.

REFERENCES

1. U.S. Nuclear Regulatory Commission. Regulatio Advancef no d Nuclear Power Plants; Statemen Policyf o t . Federal Register ,. 130 VolNo , 51 .Jul , 8 y 1986, pp. 24643-24648.

2. U.S. Department of Energy. Top-Level Regulatory Criteria for the Standard HTGR. HTGR-85-002, Rev. 1, June 1985.

. 3 Gas-Cooled Reactor Associates. Utility /User Requirement e Modulath r fo sr High Temperature Gas-Cooled Reactor Plant. GCRA 86-002, Rev , Marc1 . h 1986.

4. U.S. Nuclear Regulatory Commission. Safety Goals for the Operation of Nuclear Power Plants; Policy Statement. Federal Register, Vol. 51, No. 149, Augus 1986, t4 28044-28049. ,pp .

293 ACKNOWLEDGEMENTS

The authors would like to thank Mr, A. C. Millunzi, Acting Director, Division of HTGR, U.S. Departmen Energyf to approvar ,fo publiso t l h this paper. Thanks are also expresse managemene th o A Technologiet dG f o t s Inc permissior .fo o nt write and present this work, which was supported by the Department of Energy, San Francisco Operations Office Contract DE-AC03-84SF11963.

294 COMPARISO REGULATORF NO Y ASPECTS IN DIFFERENT COUNTRIES

K. HOFMANN Rheinisch-Westfälischer Technischer Überwachungs-Verein e.V., Essen, Federal Republi f Germanco y

Abstract

High Temperatur Cooles eGa d Reactors e (HTGRth r )fo generatio electricitf no present a e yar t only operated in the United States of America (USA) and the Federal Republic of Germany (FRG). Major development wor alss i k o eviden Hige th hn i tTemperatur e Reactor sector in these countries. Aspects of licensing practice in the USA and the FRG are therefore dealt with you to give possible impetus for the procedure to be adopted in countries where the HTGR technology will be introduced.

The design criteria and the technical codes must be supplemented with regard to the requirements of High Temperatur Cooles eGa d Reactors. Move thin si s direc- tiod experiencnan e applicatioth n i relee th -f no vant codes to High Temperature Gas Cooled Reactors, codes which are mostly oriented to Light Water Reac- tors (LWR), are to be found in the USA and the FRG in conduct of HTGR licensing procedures. Some criteria e applie b havanalogoun a o t en i d s manner e analo.Th - gous applicatio criterif no HTGRr fo a s depende th n so HTGR-design and refers mainly to the criterion for testability with respec periodio t c inspectiond san tests criterie ,th shutdowr fo a n systems, residual heat removal systems, reactor coolant boundard an y the reactor containment.

The IAEA standard "Design for Safety of Nuclear Power Plants, A Code of Practice" is applicable to HTGRs without major difficulties.

295 l. Introduction

High Temperature Gas Cooled Reactors (HTGR) generatioe th user fo d electricitf no present a e yar t only operated in the United States of America (USA) and the Federal Republic of Germany (FRG). Major development wor alss i k o evidenHTGe th R n ti secto r in these countries. It is therefore worthwhile examining aspects of licensing practice in the USA and the FRG to give some idea of the procedure which could be adopte countrien i d s wher HTGe eth R technologo t s yi be introduced.

What follows give overvien sa f licensinwo g practice in the aforementioned countries and it deals with particular regulator desigd yan n aspect Higf so h Temperature Gas Cooled Reactors.

2. Nuclear A LicensinUS e th n i g

e e federaTh th s lha governmenA US e th n ti responsibility for virtually all aspects of nuclear licensing U.Se .Th . Nuclear Regulatory Commission centralizee th (NRCs i ) d regulatore yth bodr fo y nuclear licensing process. Of particular interest are the Office of Nuclear Reactor Regulation and the Office of Standards Development, which deal with licensing technical reviews and development of technical requirements respectively.

The licensing of a nuclear power plant in the USA comprises essentially two steps: construction permit (CP) proceedings and operating license (OL) proceedings o stage tw centeree e sar . Th submittan o d l Preliminare oth f y Safety Analysis e Reporth d tan Final Safety Analysis Report for the CP and the OL proceedings respectively.

296 e fundamentaTh l statute implementin Atomie gth c e Cod th Federaf eo Titlf s Energo i 0 t 1 el yAc Regula- tions (10CFR). Parts of particular interest are: 10CFR Part 20 "Rules for Protection Against Radia- tion", 10CFR Par 0 "Licensin5 t Productiof go d nan Utilization Facilities", and 10CFR Part 100 "Reactor Site Criteria". 10CFR Par provide0 t5 basie sth c technical guidance for the design of nuclear power plants seriea n i f appenbods ,so d it botan y n -i h dices. 10CFR Part 50 Appendix A is a listing of the "General Design Criteria for Nuclear Power Plants".

The basic requirements are expanded and/or interprete othey b d r documents NRCe issuee th Th .y db NRC Regulatory Guide guidelinee sar s withoue tth force of law providing direction for meeting the regulations and may describe an acceptable solution. In orde o standardizt r e safetth e y review f licensso e applications, NRC has issued Standard Review Plans which describe the basis of NRC technical reviews. Neither Review Plan has the force of law.

A non-regulatory sourc technicaf eo l criteria for nuclear consensue plantth s i s s standards develope y industrb d y with participatio C staffNR e .th y nb These standards are usually issued by the American National Standards Institute (ANSI) after being prepared by working groups within the framework of professional societies Americae , th suc s ha n Societf yo Mechanical Engineers (ASME), the American Nuclear Society (ANS), or the Institute of Electrical and Elektronics Engineers (IEEE).

3. West German Nuclear Licensing

The centra nucleale th rol n i er licensing process is assumed by the State Supreme Licensing Authority of the individual German State. The applicant for a nuclear facility license applies to the

297 Supreme Licensing Authority of the State where the facility wil locatee lb d (fig Tha. 1) .t State then coordinate e licensinsth g proceedin ultimateld gan y issues the license. The State Supreme Licensing Autho- rities place heavy reliance upon outside experts to conduct the technical review of license applications. The supervision of operating plants is conducted by the State Supervising Authorities.

Bundesministeriu r Umweltmfü , Naturschutz und Reaktorsicherheit Inquiry BMU Inquiry Federal Ministry of Reactor Safety

Comments] Order Advlcef lor er I K Technische Gesellschafr fü t Order Reaktor- Überwachungs-Vereine Reaktorsicherheit Sicherheitskommission TÜV GRS RSK Technical Company for Reactor Inspection Agencies Reactor Safety ooO Safety Commission Preparation Information Ordei r IK I Assessmen7 t Licensing Authority General Public Instruction (Supreme State Authority)

Application lApprova r Rejectioo l n

Applicant

FIG. 1 Licensin. g proces n FRGi s .

The States do not, however, implement and enforce the Atomic Law completely autonomously. The Federal Governmen responsibilite th s tha authod yan - rity under the Atomic Law to ensure State compliance coordinato t d an w e activitieela th wite e th hth f so various States to obtain consistency. This task was assigned to the Federal Ministry responsible for reactor safet U (BundesministeBM y Umweltr fü r , Natur- schutz und Reaktorsicherheit), formerly to the Fede- ral Ministry of the Interior (BMI). The BMU nuclear staff must stil loutsidf o mak e eus e technica- lex pert mucr theifo f sho r technical work addition .I n to outside technical companies, BMU makes continuing

298 use of an advisory body of scientists and engineers very simila U.Snature n i rth .o Advisoret y Committee on Reactor Safeguards (ACRS). This advisory committee, calle Reactoe th d r Safety Commission (RSK), reviews each license applicatio wels na generis la c safety issues. Ther alss i e Radiatiooa n Protection Commission (SSK) that perform similasa r advisory functio radiation ni n protection matters.

As mentioned above, botState th h e Supreme Licensing Authority and the BMU make use of outside technical experts during license proceedings. Each of the Stateprivata s nonprofit sha ebu t companf yo technical experts called a Technical Inspection Agen- cy (TÜV) TUV'e .Th s have numerous functions fror mca inspections to boiler certification, and most have a nuclear division which- actnu experts e sa th r fo s clear licensing and supervising bodies.

Another nonprofit company Compane ,th r fo y Reactor Safety (GRS), performs technical contracts varioue foth r s TÜVs Supreme ,th e Licensing Authori- ties, the BMU, and the RSK. Other organizations having an impact on German nuclear licensing criteria include the Länderausschuss für Atomkernenergie (State Committee for Nuclear Energy), the Nuclear Safety Standards Committee (KTA), and the German Institute for Standardization (DIN). The first organizatio coordinatina s ni e Statee gth th bod d f syo an BMU, whil e latte th edeao tw rl witdevelopmene th h t of standards.

The construction permit and operating license are normally issue numbea n i dindividuaf o r l stages e for ith f so-callenmo d partial licenses. Each stage covers a part of the erection of the plant or of its operation. Wite firsth h t partial licens erecr efo - tion, the site must also have been conclusively asses- e assessmenth d e safetsean th d f yto concepa s ta

299 whole must also have been completed resultina n gi positive judgement. After the construction permits or operating licenses have been granted, erection work on site of the part of the plant concerned or the operation stage concerned can commence.

The Atomic Energy Act lays down, as an essential prerequisite for granting a license that preventive measures neede preveno t d t damage arising from erection and operation must be taken in accordance wite stat th hf scienc eo technologyd ean e .Th Radiation Protection Ordinance (Strahlenschutzverord- nung) lays down limit value systed san m requirements aframewora s k within whic handline hth radioacf go - tive substances must take place.

The Federal Minister of the Interior, - new BMU -, together with the licensing authorities of the Federal States have issued "Safety Criteria for Nuclear Power Plantsdetaio t necessare / lth /! " y preventive measures. These Criteria contain protection objectives and basic requirements for the design of systems. A set of nuclear regulations has been Nucleae th drawy b r p nu Safet y Standards Committee (KTA o specift ) detain i y e statlth sciencf eo d an e technolog e desigth realizatiod n i nan y f indivino - dual components e "KT.Th A Rules" take accoune th f to available conventional technical regulations, but generally go far beyond them.

e ReactoTh r Safety Commission (RSK draws )ha n guidelineK uRS p s specificall Pressurizer fo y d Water Reactors. These guideline morr s fa whice e detaihar - led than the safety criteria serve to unify the safety requirements.

300 4. Applicable Codes and Guides for High Tempera- turCooles eGa d Reactors

G safety-technicaFR e Inth l aim aspectd san s for plant and environment are laid down in the "Safety Criteri Nuclear fo a r Power Plants" /!/. These apply strictly to the Light Water Reactors and also to other reactor types suc HTGRs ha indirectld ,an y when considering plant-specific systems.

The analogous application for HTGRs refers criterioe mainl th testabilite th o t yr fo n y with respec periodio t c inspection testsd san crie ,th - teria for shutdown systems, residual heat removal systems, reactor coolant boundary and the reactor containment orden .I avoio rt d difficulties which arose from analogous applicatio licensine th n ni g procedur THTR-300e th f eo ,drafa safetf to y criteria was drawn up for HTGRs /2/ and the Safety Criteria for Nuclear Power Plants were redrafted /3/ to include the HTGR.

Planning specifications /4/ interpreting the Safety Criteria for Nuclear Power Plants for the THTR-300 have been especially adapte workina s a d g basi r thosfo s e involve e licensinth n i d g procedure of THTR-300. Experienc bees eha n positive with this procedure .A simila r procedur e plannes ei th r fo d licensin HTR-500f go desige th , whict na s stagehi .

Rules issued by the KTA, which apply either fully or analogously to HTGRs including rules especially developed for HTGR design, have been re- e detaileth r ferrefo d o t ddesig HTGf no R nuclear power plants, such as HTR-Module and HTR-500. Rele- vant item HTGRt includen so ye st n whicno i d e har nuclear standard e laisar d dow specificationn i n s agreed within the licensing procedure and which can be orientate conventionao t d l standardsN DI suc s a h

301 (Deutsches Institut für Normen) or AD (Arbeitsgemeinschaft Druckbehälter). Foreign codes, such as the ASME code, are also sometimes referred to.

KTA wor HTGr bees fo kRha n intensified within the las yearo speciaa tw t y sb l HTGR-Subcommittee (HTR-Unterausschuss) in KTA. Also there are major activities on the part of the KFA Nuclear Facility at Jülich together with other companie proposinn si g technical rule HTGRr fo s . They involve proposalr fo s rules for high temperature materials and components for both electricity-generating and nuclear heat- generating HTGR power plants. Temperatures greater than 800° includede Car .

Activitie developinn i s g criteri guided aan s for HTGRs have alsoe th bee n i n undertakeA US e th n i n past, e.g.: drafa f "Nucleato - r Safety Criterir fo a the Design of Stationary Gas Cooled Reactor Plants" with supplements prepared Americae bth y n Nuclear Society /5/,

an analysis showing whether the Regulatory Guide e applicablsar o t e HTGRs,

revision of the "General Design Criteria for Nuclear Power Plants" presenten i d Appendi Parf Titl, o Cod, t50 xA 10 f eeo Federal Regulations applicatioe th r fo , n to HTGRs (draft) /6/,

expandin ASMe gth E cod respecn i e t to high temperature design.

Sinc relevane eth tA codeUS guided e san th n si were draw againsp nu backgroune tth Lighf do t Water Reactors, it is necessary there as in the FRG to make

302 certain interpretations for HTGRs, and the drafting of special codes and guides for HTGRs might be the objective for*the future. Worthy of mentioning here are the existing supplements to the ASME code, which are also referred to in the FRG with regard to design.

Eighte Oth n h International Conference th n eo HTR Septemben i G r this yea heare rw d curs i tha t -i t rently planned by NRC to review the Preliminary Safe- ty Information Document (PSID) for the "Advanced HTGR" 111. The main purpose of the review will be to establish the licensing criteria which apply to the advance makdo t assessmenn HTGea d Ran potene th f to - tial of the proposed design to meet those criteria. The licensing criteria will be established by building upon existin criteriaR gLW , where applicable, and by developing additional criteria, as necessary to addres unique th s e aspectHTGRe th .f so

5. HTGR-Specific Features

unique Th e feature HTGe th R f musso e tb discussed during regulatory work, important items here are:

5.1 Core Desig Systemd nan Contror fo s d lan Shutdown of the Reactor ä

Similarly to LWRs, two independent and diverse reactivity control systems are required. One of these shal capable lb o shutdowt e reactoe th n r frol mal operational and accident conditions for a sufficient period; the second shall be capable of maintaining cold shutdown for an unlimited time.

The following inherent safety characteristics of the HTGR

- graphite structure with a high heat capacity, thermal conductivity and phase stability

303 - phase stability of the coolant helium

fuel elemen sensitivt tno overheatino et g

negative temperature coefficien reactivitf to y

should be considered when specifying requirements for shutting down the reactor.

Therefore, a design is acceptable in which hot shutdown conditions are provided by increasing the average core temperature. This has been verified experimentally with the AVR High Temperature Reactor Power Plant in the FRG by turning off the helium circulators without movin controe gth l rods which results in a reduction of the coolant flow and an limited increase of the core temperature /8/.

5.2 Reactor Coolant Boundary and Pressure Bearing Vessel

case ILWR th f nesmalo d san l HTGRs reace ,th - tor coolant pressure boundary includin pressure gth e vessel consists entirel metallif o y c components.

Accordin desige th HTR-50f o no t g s i t 0i preferable to specify requirements for the metallic components which represent the enclosure of the reactor coolant and for the pressure bearing vessel sepa- rately.

In detail reactoe ,th r coolant boundary consists of

the leak-tigh tpressure lineth f ro e bearing vessel,

304 - penetrations throug vessee hth l including their closures,

reactor coolant piping including the first isolation valves,

pipelines penetratin vessee gth d lan interfacing with the reactor coolant at their outer surface (e.g. heat exchangers primare inth y circuit).

In addition to general requirements to design, testing, material leakagd san e monitoring instrumentation, HTGR-specific considerede featureb o t e sar , e.g.penetratione th : pressure th f so e bearing vessel must be secured against outward forces, and consequences of closure failure must be mitigated by limi- tinblowdowe gth n flow area, e.g provisioy .b f no flow restrictors.

5.3 Residual Heat Removal Systems

Reliable residual heat removal systems are require operationar fo d accidend lan t conditions.

e followinTh g feature sconsideree b hav o et d when establishing requirement accidene th r sfo t residual heat removal system:

No total loss of coolant occurs in an HTGR-system that a minimum helium pressure remainprimare th n si y circuit.

Ingress of foreign media, e.g. water or air into the primary circuit or chemical reactions may occur.

305 The inherent safety characteristics shoul considerede db , discussed earlier, e.g. the properties of graphite and the fuel element.

The proposal for the design requirements includ aspecte th e , tha inherenf i t t characteristics can assure residual heat removal or storage after accident thao s t design exceededt limitno e sar , hardware requirements can be softened, e.g. the requirement for meeting the single failure criterion with respecresiduae th o t l heat removal system durin maintenancs it g e suspendeb aden n a e ca - r fo d quate period.

A three hour interruption of residual heat removal has been investigated for the THTR-300 power plant /9/. Accordin thiso gt e structur,th e th f eo cor preserves i e orden i d r tha residuae tth l heat removal can be resumed by the active systems after three hour thad reactoe broughe san th tb n rca t into a safe state without exceeding radiological limits at all times.

If development and engineering lead to a design where heat removainterruptee b y lma lona r g fo dtim e period without doing harm to the public, so-called "passive safety obtaineds i " . Progres than si t directio evidenAdvancee s i nth r fo td HTGR, HTR-Moduld ean HTR-500.

5.4 Nuclear Reactor Containment

In the design requirements of the containment a high-pressure containment or a controlled vented j containmen adequatee tar shoult .I statee b d d that during external events

306 the containment shall remain both leak-tigh d structuralltan y intactt i f ,i canno showe tb n tharequiremente tth s of the Radiation Protection Ordi- nance /10/ with respect to radioactive releases is adhered to as a result of acciden operationar to l leakage froa m non-leak-tight structurer ,o

the containment shall only remain structurall e showb yn nintactca t i f ,i that even without leak-tightness the requirements of the Radiation Protection Ordinance are met.

6. Applicability of the IAEA Standard "Design for Safety of Nuclear Power Plants, A Code of Practice" to HTGRs

On the international level, the Code of Practice "Desig Safetr fo n Nucleaf o y r Power Plants" /ll/ is published within the IAEA Safety Standards. This code can be adopted by countries intereste nuclean i d r energy discussione .Th o t s ,a whether this is applicable to HTGRs without detailed interpretatio restrictee b n nca o fivdt e items- ,be caus othee th e r e standarpartth f so d cite fundamental protection objective contair so n plant non-specific requirements.

6.1 Provisions for In-Service Testing, Maintenance, Repair, Inspection and Monitoring (Sectio /ll/f o 9 ) n2.

In principle, here measures are required for in-service testing, maintenance, repair, inspection and monitorin e functionath f go l capabilit compof yo - nents HTGe th r somR . Fo f componenteo specian si l

307 design a restricted accessibility applies. In section 2.9 of /ll/ this restricted accessibility is taken into account by requiring "adequate safety precau- tions" to compensate for potential undiscovered failures,

6.2 Reactor Core (Section 4 of /ll/)

The criteri corr d fuefo aean l desigd nan reactor contro d shutdowlan n system layou kepe tar t sufficiently genera thao ls appliete b then yca d also for HTGR power plants.

6.3 Reactor Coolant System (Sectio f /ll/o 6 n )

comprise/ I I Sectio/ f o s6 n design requirements for the residual heat removal systems and for the reactor coolant boundary including the pressure vessel. The requirements for the reactor coolant boundary are general so that they can be applied to HTGRs as well. However criterir ,fo a more specifio ct HTGRs, it is better to differentiate between the enclosure of the coolant and the pressure bearing vessel for some HTR-designs as has been discussed above.

Section 6.6 of /ll/ "Emergency Core Cooling" is LWR-specific in essential aspects so that interpretations for the application to HTGRs are necessary. Thus, sectio 6 require6. n emergencn sa y core cooling system for the case of the loss-of-coolant accident in orde complo rt y wit desige hth n clad e valuth -f eo ding temperatur e fueth l f elementso e , whics hi LWR-specific.

4 Containmen6. t System (Sectio /ll/f o n8 )

Withi requiremente nth sectiof f /IIso o 1 /n8. "Purpose of Containment System" it is possible to

308 hav ventea e hermeticalla r o d y sealed containment dependin othen go r mean f limitinso release gth f eo radioactive substances. These feature consistene sar t with the requirements for an HTGR-containment within Chapter 5 of this paper. Effects of potential HTGR- specific energy sourcecontainmene th n so t structure, e.g. from reaction witr ai h f so graphit fore th - r eo matio combustiblf no e gases e include,ar d within section 8.2 of /ll/ "Containment Structure Strength".

5 6. Conclun sio

The conclusion to be drawn is that the IAEA standard "Design for Safety of Nuclear Power Plants, A Code of Practice" is applicable to HTGRs without major difficulties.

References

/!/ Der Bundesminister des Innern, Sicherheitskriterien für Kernkraftwerke Bonn 1977

/2/ TUV-Arbeitsgemeinschaft Kerntechnik West, Sicherheitskriterie Anlager fü n r Ernergiezu n - erzeugung mit gasgekühlten Hochtemperaturre- aktoren Draft September 1980

/3/ Der Bundesminister des Innern, Sicherheitskriterien für Kernkraftwerke y 198DrafMa 41 2 t d BBC/HRun G HK /4B / Planungsgrundlagen für die Errichtung des 300 MWe-THTR-Prototyp-Kernkraftwerks Hamm/ Uentrop : 22.06.197of s a 6

309 /5/ American Nuclear Society, Nuclear Safety Criteri Desige th f r no afo Stationary Gas Cooled Reactor Plants ANS-23 Subcommittee, Rev , Draf9 . ., 2 tNo Illionois (Jan. 1974)

/6/ Code of Federal Regulations 10 CFR Part 50, Draft Appendi Genera, A x l Design Criterir fo a High Temperature Gas Cooled Reactors, US Government Printing Office, Washington (1975)

111 King, T.L al.t .e , NRC-Review of the Advanced HTGR - A Status Update and Perspective Eighth Annual International Conference HTGRe o th n Diego ,Sa , Sept. 15 and 16, 1986

/8/ Knüfer, H., Abschaltvorgänge beim AVR-Hochtemperatur- reaktor Brennstof-Wärme-Kraft 26 (1974) Nr. 12, Pages 495 - 502

/9/ Kietzer al.t e , ,,K. Safety Analysis of the THTR-300 MW(e) - Prototype-Reactor and Future HTRs under extreme Accident Conditions International Conference on Nuclear Power and its Fuel Cycle, Salzburg (1977)

/10/ Der Bundesminister des Innern, Verordnung üben Schutrde r Schädezvo n durch ionisierende Strahlen (Strahlenschutz- verordn.-StrISchV) Bonn (1976/1982)

/ll/ IAEA, Design for Safety of Nuclear Power Plants ,A cod f Practico e e Vienna (1978)

310 GAS-COOLED REACTOR APPLICATIONS (Sessio) nE HTGR APPLICATIONS, PROSPECTIVES AND FUTURE DEVELOPMENT

H. BARNERT Institut für Reaktorentwicklung, Kernforschungsanlage Jülich GmbH, Julien, Federal Republic of Germany

Abstract

The high temperature reactor (HTR) is an advanced reactor concept s "unrestrictedli t i : y safe against catastrophes", can be used for electricity production and process heat application economicalls i d an s y attractive offert I . broaa s d variet f possiblo y e applications in the future energy market. The next e developmenth ste n i pe system th s i ts demonstration, e.g. with AVR-extension.

1. HTR for world energy supply Nuclear energy is established in the market for electricity production in many countries of the world. The main driving force is its economical competetiveness. The high temperature reactor (HTR) is considered to be an advanced reactor concept. Of advance are the followin R offerHT o factse stw gth "unrestricte) 1 : d safety against catastrophes" and 2) it opens up a broad variet f applicationo y addition i s o electrit n - city production: thes procese ear s heat applications. These applications will be described in the following from the development point of view. The users prospectives in electricity production and in coal refinement are described by Dr. Klusmann in this meeting, Session G, /!/.

2. Status of research, development and demonstration The status of the work on research, development and demonstration, done since more than two decades on describee e HTRb th n ,ca s followsa d , FIG. 1 .

313 statuD D+ + sR R Hf electricity process heat•

nettstyi : (MR

FIG. 1: Status of the Research and Development and Demonstration Work on High Temperature Reactors

For HTR electricity production the R+D+D-status has reached tha f Commerzializationo t . For HTR process heat applications the R+D+D-status has reached tha f semi-technicao t d pilolan t plant demonstratio maie th n f componentsno nexe .Th t steps in R+D+D requir e systemth e s demonstration, e.ga .vi e extensioth e AVRth . f o n

3. Unrestricted safety against catastrophes e accidentTh Chernobyn i s Harrisburd an l g have influenced and may further on influence the development and application of nuclear energy in the world to a large extend e analysi.Th s indicates quali w thane ta - n reactoti y r e necessarysafetb y ma y . Thi w qualite summarizeb sne y ma y s followsa d : catastrophic accidents should be impossible. The R+D work on the HTR since Harrisburg and discussions after Chernoby n advancelo d technologn i y nuclear energy, /2/, can be summarized as follows, e HTR-systeTh FIG: 2 . m provides e developmenth y :b t and the application of "inherent stabilizing qualities", e.g. the negative temperature coefficient of the reactivit d decaan y y heat remova y thermab l l radiation and conduction: "Unrestricted safety against catastrophes, even if mistakes are made by operators".

314 th€ HfR system provides :

inherent stabiliiing qualities unnestricted safety against catastrophs even if mistakes are made by operators .

FIG. 2: Fundamental Safety Qualities of the HTR-system: unrestricted safety against catastrophs

e developmenTh applicatiod an t f inhereno n t stabilizing qualities have the potential to produce, FIG 3, n enhancemena 1) e samf basito th ey cb safet- d an y effects - 2) a reduction of investment costs by a decrease of control systems, additional systems and auxiliary systems .n indicatioa Thi s e statei sth r -fo n ment that economics and safety do not necessarily represent conflicting goals.

inherent stabilising qualities

1) enhancement f basico safety tavAt.fa. zfttzkAM OMA. - — 2) reduction of investment costs

FIG. 3: Effects of Inherent Stabilizing Qualities: economic necessarilt d safetno san e yar y conflicting goals

4. Economical competetiveness of HTR

4.1 HTR electricity production The results of recent studies by Arbeitsgemein- schaft Hochtemperaturreaktor,/3/ and /4/, can be summarize s followsa d , FIG: 4 .

315 electr. production , economics

evaluated. co«di'tiov.s )

1 competitive » Hard COOl HTR mean SIM., Smatt si'j

2. competitive » DV/R J250 (equaj to l S,-K World Market Coat u™

FIG. 4: Statments on the Economics of HTR Electricity Production

In the field of the production of electricity the HTR (medium and small size) is competetive compared to electricity on the basis of German hard coal in base R (meaHT e n loadth size d ,an , single uni multiplr o t e small units competetivs i ) about (a e same tth e costs) compared to electricity from large PWRs, german condi- potentialls i R HT e tionsth y d competetiv,an e compared to electricity from world market coal, base load. e fossiTh l competitor stakee b coursf hav o o nt - e - e including additional investment for environmental reasons. For indicating the cost situation for HTR electricity production the "cost-flow-diagram", following /3/, is give t indicaten FIGI i n . .5 e followingth s : from fuel elements for about 1 $/GJ high temperature helium is produce HTR-core th n i dcostt a e f abouo s 3 $/GJ3. t , HfR electr production.economics

Steam furbo- ^ Core * Gm Kachinc Fuel High -Temp. Live Electric HeUum Steam 33 __ % G3(Ft) ffj(Ke) (is)

FIG Cost-Flow-Diagra: 5 . ElectricitR HT r mfo y Production

316 which than is converted in life steam in the steam generato givo t r e abou 4 $/GJ4. t , which finall cons i y - verted in the turbomachine into electricity at about 13.6 $/GJ, respectivel jzi/kWh5 y e ("real" cost evaluation method DM/$)5 ,2. . procesR HT 2 s4. hea generan i t l market Frocost-flow-diagrae mth HTR-electricityr fo m - production it can be concluded, that HTR-heat in the form of high temperature helium (primary helium circuit) costs abou 3 $/GJ3. t , FIG. 6

Process Heat Cost ftgures

HfR Coat (Autothermot)

HigK Temp. Hand World /muittas\ Harret Ueo.CJ (KG) CbaL \ ' ' (frinary He) 1 6.3 -61-- -'•^33 4-.? Coot Î.6 2.0 -(.? -l.f Kemt 1

FIG Compariso: .6 Cosf o n t Figure r Procesfo s s Heat: HTR process hea d procestan s heat from coal

In autothermal conversion and refinement processes e requireth d process hea takes i t n froe fossith m l input. In most cases this heat production consists of a partial combustion of the fossil feed with oxygen. A rough calculatio r coalfo n particulan ,i r gerfo r - man hard coal respectively world market coal, FIG. 6, indicates, that process heat from these sources costs about 6.3 respectively 4.7 $/GJ, /5/. e rougTh h calculatio fossie th coste r th l fo sf o n process heat include the costs of the fossil energy carrier, FIG. 7, costs for oxygen, separated from air, and cost r equipmentfo s . Right know, autumn 1986, the market prices in Germany for natural gas and oil are about the same as the world market price for coal.

317 Primary Energy Carriers cost

German Hard CbaC 3.6 Z60 DfVt V/odd Market CbaC 2. €b i/t

(rerrviaM. uofru-te. 1.9 t fe6Î7W/ w-

OIL (import] (Hi'd W/t«i

Oit steaw. injection 3.6 -SA iOo ^/barrt-i l CO^ , Chawx'caCs 4.5-7.2

Oil sands/ shales 6.\-l8 11-100 li

LWCr 6'wporfc ) SAU - 30-fe V

FIG Cos: 7 . t Figure n Primarso y Energy Carrierr fo s the Federal Republi f German o cd 198 mi 6n i y and Future Prospects

Therefore, the cost evaluation for fossil process hea abous tha same tth e result FIGe ,se .. 7 In summary it can be concluded that the HTR-process is by a factor of about 2 competetive compared to German hard coal and by a factor of 1.4 competetive compared to world market coal as well as natural gas and oil. An overview on cost figures of primary energy carriers is given in FIG 7. In the upper part it describes cost figures for primary energy carriers in autumn 1986 for the Federal Republic of Germany and Western Europe. In the lower part it describes cost figure r futurfo s e primary energy source s thessa e may be of relevance for Western Europe. e cosTh t situatio energe th n yo n marke characs i t - terizee sharth py b d decreas prica f ef oilo o e t i : costs now about 10 $/barrel, equivalent 1.8 $/GJ. A few years ago it has been more expensive by almost the factor of 4. For the future increases of the oil price e expectedar n indicatio.A cose gives th i nt y b nfigu - res of enhanced oil recovery and even oil production from oil sands and oil shales, FIG. 7.

318 An overview on the broad variety of possible appli- given i e firsFIGn R i nTh HT cation. t8 e brancth f so h is HTR-electricity production by the turbine cycle and in cogeneratio wels a n distric s la t heat e secon.Th d branc calles i h d "process heat applicationss i d an " devided into three categories: 1) HTR process heat applicaton for recovery, conver- siod refinemennan f fossito l energy carriers, 2) HTR process heat applications via the chemical heat loop (EVA/ADAM-system) and 3) HTR process heat applications via the thermo-chemical cycle for the production of hydrogen and oxygen from water.

Turbine Cycle iü,ct*ic*h£^ CogenerattOM.

u _ • i LjT"D«b\AOMiUdin (hü. H l KT1 s*™ >j«*o^ ^ Javic

~~ChemicaC Heat U»p .. , .. . ? Thermo-CheiKicaC G/de~~

~~FIG. 8: Broad Variety of Applications of the HTR in Electricity Production and in Process Heat Applications~~

followine Ith n g three example procesr fo s s heat applications in combination with fossil energy carriers are given in overview. The R+D+D-work on the technology for this application is performed (and has been performed) in the projects "Prototype Plant Nuclear Process Heat (PNP)" and "Nuclear Long Distance Energy (NFE)". The results of this work are described by Dr. Klusmann, /1/.

4.3 Integratio coalf no , stee d nuclealan r energy The energy conversion system "Integration of coal, steel and nuclear energy" is a concept for the production of methanol (energy alcohol) and pig iron on

319 the basis of German hard coal, iron ore ( + additional substances) and nuclear energy in the form of process heat from the HTR. These concepts combine three groups of processes, FIG. 9, in particular: e stea1th ) m coal gasificatio partian i n l gasification for the production of gas and coke, 2) the Klöckner-steel-gas-process for the reduction of iron ore in a converter for the production of pig iron and converter gas and e jointh t) 3 applicatio produce th f no t gases from the gasifieconvertee e productioth th d r an r fo r n of methanol.

~~(nfcegrgtu f Coatwo , Steel cmct~~

~~Hard Hz,a Hetfcmol Coal —— > CoaC, ^ Gas —— ^ HeUianoL Conçu' U — > Mm -) •tfoning ——> • Î Cok< Cbnvisrkrqcu aiV •I CO Klôckner V.Steet 'Gas — * Slag~~

~~FIG. 9: Simplified Flow-Diagram for the System "Inte- gration of Coal, Steel and HTR" for the Pro- ductio e Basif Methano o th ng Iro Pi n so nd an l of Hard Coal~~

An economical evaulation has indicated that the system "Integratio coalf o n , stee d nuclealan r energy" (in german: Verbund von Kohle, Stahl und Kernenergie) has the potential to be economically competetive compared with the market situation of 1984/85, FIG. 10. The integration can convert German hard coal at about t 10(257,$/ 0 - DM/tg pi ) d intan t o $/ methano 0 20 t la iro t 150,na - (respectively 167,-) $/t, including (ex- cluding )a subsid cokn o y e (Kokskohlenbeihilfe) given for blast furnace operation. f intereso s i motos t ta $/ Methano r0 fue20 t lla and as heating fuel. As motor fuel methanol in 1984/85 seemed to be economically competetive compared to

~~320 Integratio Coat- nof , Stee d 4fTfan l e~~

~~Integrated 15) $/frt0(w o Iron «L VOL.. SubtidSubri*j fob.)~~

~~c/>/> nuK* CooL « j. /, i 2i vA/ */tiu lT ia 1^ r2.SfîH/» t j v l^f\ Hi/^- — — — — — — — — . — _ t MO i'c/aNJVT. ü'w'tcokuc —>^• -IAI*KT- Blast fixrnace_ 150-160 $/t '~~

~~FIGCompariso: .10 Integratee th f no d System versus tht Integrateeno d Integratioe Systeth n i m n of Coal,R SteeHT d lan~~

high octan gaslin Otto-carsn i e . Meanwhill oi e eth pric thereford ean gasoline eth e pric decreaseds eha . The results of the economic evaluation can therefore be summarize followss a d : economical attractiveness seems to be achievable at a price level for oil of abou $/barre5 2 t l (an 5 DM/Sd2. ) that means compared to oil from the North Sea, future Norwegian Shelf, /6/,

4.4 In-situ oil sand synthetic crude production Another promising example for the application of process heat from nuclear energy, in particular HTR- process e in-sitheatth s l sand,i uoi s synthetic crude production l sanreference oi . Th d e mininth s i eg project "Syncrude Forn i " t McMurrey, Canada operation ,i n yearw sincfe sea with production cost abou/ f s$ o 9 1 t barrel (1984). For the future it is important that most of the oil sand is deeply burried. Therefore "in-situ"-production has been proposed, e.g. by steam injection. requiree Th d injection producee b stea n mca y b d the HTR. Compared to other nuclear reactors the HTR producn ca e high pressure stea - almosr re fo ml tal quirements feasible. In additio o injectiont alsn ca o nR HT stea e mth deliver hearefinemenr fo t t processe electricitd san y

321 for the production of hydrogen from water. Following these lines increasing amount f nucleao s r energn ca y e integrateb d inte recoveryth o , conversio refined an n - e overalth men f to l process. In a feasibility study, /7/, on the basis of the existing plan Forn i t t McMurrey r in-situ-profo t ,bu - duction by steam injection it has been shown, that the product yield could be increased from less than 70 % stepwise upto more than 130 %. At the same time the side production of CO» is reduced equivalent ly. e fundamentaTh l process step describee sar n i d FIG. 11. The HTR produces the injection steam, as well as process stea d electricityman conversioe .Th n consist upgradingf so , flexicokin d hydrotreatingan g e .Th final products are synthetic crude, methanol and purge coke.

~~n-sfcu oilsqnot synthetic crucU.~~

~~nthetic Crude Hettoiol u~~

~~Simplifie: 11 FIG. d Flow-Diagra e In-sitth l r uOi fo m Sand Synthetic Crude Productio Proe th - r fo n duction of Synthetic Crude and Methanol using HTR Process Heat~~

4.5 Natural gas conversion to liquids Naturas toda i d wil s an ye morlb ga ln futur i e e an important energy carrieworlde th t todan .i Bu r y s exporteii t d n extena onl o t f abouyo d compare% 15 t d productione th o t e reaso.Th obviousls i n expene th y - sive transportation by pipe lines or as liquified natural gas. Therefor e conversioth e naturaf no s intlga o liquids, e.g. methanol (ma togethee yb r with ammoniaa s i ) means to make it a "world trade energy carrier".

322 This conversion can be done by autothermal pro- applicatioe th cessey b r so HTR-procesy nb s heate .Th latte woule on r d increas produce eth t yield from about 60 % to about 110 %. An overview on the process for the synthesis of methanol from natural gas (together with carbon dioxide) is given FIG. 12.

~~Methanol- Synthesis Mfr (n«cot)~~

~~ln,«,„.„. Natura Methan HeWlCWO1 f ^ lT t~~

~~vGas~~

~~Simplifie: FIG12 . d Flow-Diagra Conversioe th r mfo n of Natura s intlGa o Methanol usin R ProcesgHT s Heat~~

An analysi markee th n tso possibilitie r remotfo s e fields of natural gas of limited content, /8/, shows the following results remotr :fo e natura fields lga s of limited conten conversioe th t n into "liquids" (methanol + ammonia) using HTR-process heat may be economically attractive, even if the production costs are higher compared to the autothermal process, because the increased yield produces a larger cass flow for the operator.

5. AVR-extension As indicated in FIG. 1, the next step in the R+ D+D-work for process heat applications is the systems demonstratio coupline th f no higf go h temperature helium fro nucleama r reactoheae th t o t consuminr g components. Therefore it has been proposed by the d KFotherAan s thaexistine th t g AVR-plant shoule b d usethir fo ds system demonstratio extensionn a y nb . The AVR-reactor is (and has been) used for the productio steaf no m which finall convertes i y d into

323 electricity which is given to the grid. The outlet temperatur reactoe th f eo r coolan. °C t heliu0 95 s i m The proposal "AVR-extension", /9/ /10/d ,an , contains the following, FIG. 13: half of thermal power (and - de heliu e th museC r ° mas fo d0 s95 flowt a s i ) monstration of the coupling of nuclear helium to heat consuming components. These are the steam reformer (steam-methane-reforming) steam-coal-gas-generae ,th - r (steam-coal-gasificationto e intermediatth d )an e heat exchanger.

~~AYR for NUCLEAR PROCESS HEAT~~

~~Process Heat W LooM 5 p2~~

~~C ° duc s 0 ga 95 t t Ho~~

~~Heat consuming components • Steam reformer • Gas generator • Intermediate heat exchanger Confinement~~

~~KFA In co operation win AVR, HRB, Mteratom~~

~~FIG. 13: AVR-Extension: AVR for Nuclear Process Heat, Systems Demonstratio f Heano t Consuming Com- ponents~~

. 6 Future R+D-prospectives Future prospectives for research and development e "bettewor th e openey ar kb rp u dutilizatio e th f no temperature potential of the HTR". This potential will become applicabl f i eR+D-wor ceramin o k c materials, n particulai ceramin o r c heat exchangers, wile b l successful. Today the process temperature in the heat consuming processes is limited by the metallic heat transfer surfaces to about 900 °C. Both fundamental process heat reactions, the "steam-methane-reforming"-reaction and e "steam-coal-gasification"-reactioftth , woul e posib d -

324 tively influenced by an increase of 50 K - 100 K (or more) because of the following reasons: 1) better fitting of the HTR-heat-vector and 2) higher space velocit particulan (i y coar fo rl gasification) and 3) higher reaction degree (with respece requireth o t d products). The specified maximum fuel temperature of the THTR fuel is 1250 °C. Between the specified process d 125an C therC a ° 0° s i e0 temperature85 o t 0 80 f so large potentia r improvemenlfo processese th f to , particularly with respec o economicst , FIG. .14

~~HTR Temperatures~~

~~1150 T (lax PueL~~

~~T Gertxfvucs 1000 H _ '/f't/rr~~

~~700-1~~

~~: FIGTemperatur14 . e Value n HTi s R Process Heat Appli cations e betteth , r utilizatio e Temth f -no perature Potential of the HTR~~

7. Results of this report e HTTh 1R) R+D+D-wor s reacheha k e statth d f commero e - cialization for electricity production respectively e statth f semi-technicaeo pilod an l t plant demonstratio f processeo n r procesfo s s heat applications. For the latter the next step is the systems demonstration, e.g. by AVR-extension. 2) The HTR provides by the development and application of inherent stabilizing qualities "unrestricted safety against catastrophes", even if mistakes are made by operators.

325 3) This developmen te sam th offere y effectb s- s- a reduction of investment costs. Economics and safety do not need to be conflicting goals. e economiTh ) 4 ce b competetivenes n ca R HT e th f so characterized as follows: in electricity production R (mediuHT e d smalth man l size competetivs )i e compare Germao t d n hard coal, base th load d an ; HTR (medium size, single uni multiplr o t e small units competetivs )i e (abou e samth te costs) compared to large PWR, germas ni conditionsR HT e th d ;an potentially competetive compared to world market coal, base load. 5) Fro mconcludes i thi t i s d tha procesn i t s heat applications the HTR (medium and small size) is potentially competetive compare Germao t d n hard coal and world market coal (as well as oil and natural gas) procesR .HT s heat costs / abou$ 3 t3. s therefori d an a facto , y e b GJ f abouo r res2 t - pectively 1.4 cheaper than fossil competitors. However markee ,th t situatio s changeo nha t e du d the sharp decrease of oil prices. R offerHT e a ver6s)Th y broad variet f possiblo y e applications in electricity production and process heat applications. Three typical example dese ar s - e followingcribeth n i d : e syste7)Th m "Integratio coalf o n , stee d nuclealan r energy" for the production of methanol and pig iron, using HTR process heat has the potential to be economically competetive compare markee th o t d situation of 1984/85. Economically attractiveness seems to be achievable at an oil price level of 25 $/barrel (and 2.5 DM/$) that means compared to l froNorte oi th m h Sea, future Norwegian Shelf. e "in-sitth n l I san oi u) d 8 syncrude production" (réf. syncrude production in Fort McMurray, Canada, in operation since a few years) the application of HTR-process heat offers the possibility of an increase producth f eo t yield froo mt abou% 0 7 t (stepwise) more than 130 %.

326 converte9e )b Naturan ca s d lga into "liquidsd an " thereby become a "world trade energy carrier". For remote fields of limited content the conversion usin R procesgHT seconomicalle b hea y tma y attractive producte , th eve f i n s cost highee sar r compared to the autothermal process, because the increase of the product yield produces a larger cash flow. 10)The next step in the R+D+D-work for process heat applications is the systems demonstration of the coupling between high temperature helium from a reactor and heat consuming components. Therefore, s beethha enR extensioproposedAV e th f o n. 11)Future R+D-prospectives are opend up by "the better utilizatio e temperaturth f o n e potentiae th f lo HTR". This becomes feasible by ceramic materials for heat exchangers.

~~REFERENCES~~

/1/ KLUSMANN , "Futur,A. e Application HTGRe th "f o s IAEA Technical Committee Meeting on Gas-Cooled Reactor Theid an s r Applications, Jülich, 20-23 October 1986, paper no. G1. /2/ BARNERT, H., "Fortgeschrittene Kernenergietechnik" Vortragsveranstaltun Arbeitsgemeinschafr de g t der Großforschungeinrichtungen: Tschernobly und die Zukunft der Energieversorgung in der Bundesre- publik Deutschland Godesbergd ,Ba , 17./18.Sept.86. /3/ Arbeitsgemeinschaft Hochtemperaturreaktor (AHR), "HTR 500-Vorprojektuntersuchung", Zusammenfassung, Juli 1984. /4/ Arbeitsgemeinschaft Hochtemperaturreaktor (AHR), "HTR-Modul-Vorprojektuntersuchung", Zusammenfassung, November 1984. Potentiae th / BARNERT e Improvemenn /5 th "O f lo , ,H. t oCompetetivenesa f Coaf so l Refinement Proccess by Application of HTR Process Heat", HTR Economics and Market Working Party, Brussels, Dec 1985, .3 . /6/ BARNERT, H., "Der Verbund von Kohle, Stahl und Kernenergie Verbundsystes ,da m "Wasserdampf-Kohle- Vergasung mit HTR-Wärme und Klöckner-Stahl-Gas- Verfahren", Vorteile und Probleme", JÜL-2085, Sept. 198 n preparation)5(i .

327" /?/ BARNERT, H., "Oil Sand Synfuel Production using Nuclear Energy", JÜL-1956, Oct. 1984. / BARNERT/8 , "Synthesega,H. Erdgas au sHTR-Wärme"t smi , Arbeitsseminar: Erdö Kernforr Erdgad lde un n i s - schungsanlage Jülich, KFA Jülich, 1./2. Juli 1986. /9/ Kernforschungsanlage Jülich GmbH (KFA), "Umbau des AVR-Reaktors zu einer Prozeßwärmeanlage", Ergebniss Vorplanungsphase,r ede ' Juni 1985.

/10/SINGH, J., BARNERT, H., HOHN, H., ROMES, G.,(KFA), METZNER, U., CHAKEABKARABARTI, S., KÜHN, H. (Mannesmann Anlagenbau), "Untersuchun Ausler zu g - gunr primärede g n Kühlgasleitunge AVRe di -r fü n Erweiterung", Jahrestagung Kerntechnik, Aachen, S.423, 1986.

~~328 PERSPECTIVE HTGRF SO CHEMICAN sI D LAN IRON AND STEEL INDUSTRIES OF JAPAN~~

K. TSURUOKA Kawasaki Steel Technical Research Corporation, Tokyo . KATAMINT E Sumitomo Chemical Company Ltd, Tokyo T. MIYASUGI, R. ARAKI Chiyoda Chemical Engineerind gan Construction Company Ltd, Tsurumi-ku, Yokohama Japan

~~Abstract~~

HigA h Temperature Gas-Cooled Reactor (HTGR solvn )e ca th e problems in improving the economy and broadening the field of utilizing nuclear energy while maintainins g it safety o t e ,du specific characteristics of high temperature heat, inherent safety, and high burn-up. Considering the importance to use nuclear energy in non-electric fields in the future, this paper present recensa tperspective studth e n th HTGe yo th n Ri f eo chemical and iron & steel industries of Japan, conducted by the Japan Atomic Industrial Forum.

Although the heat generated by the HTGR will be expensive for use by the chemical and iron & steel industries in the rest of the 20th century, the chemical industry has many good processes which can apply nuclear heat when the cost is competitive compared with fossile fuels. Among them, the proces hydroger sfo n production coul consideree db major fo d r application iroe Th steen.& lpotentiag industrbi a r s lyfo ha usreducinf eo produces gga nucleay b d r heat with sufficiently competitive cost.

1. TEMPERATURE RANGES AND HEAT QUANTITIES REQUIRED FOR PROCESS HEAT APPLICATIONS The temperature ranges and the heat quantities required for typical processe petroleun si m refinin petrochemistrd gan y are shown in TABLE I and TABLE II, respectively.

~~329 U) oCO~~

~~TABLE I Temperature Ranges and Heat Quantities Required of Petroleum Refining Processes~~

Required Energy Corresponding Thermal Output of Process (Per Feedstock lOOObbl) Representative HTR (MWt) (Effectiveness 100%) Process Feedstock Electric Plant Capacity Temp.'C Fuel Steam (10 bbl/SD) Equivalent Steam & (10 Kcal) (ton) Power (KWH) of Fuel Electric Power Atomospheric Crude Distillation Oil 250-350 17 24 22 30 210 210

~~Vacuum Atomospheric Distillation Residue 350-430 20 20 20 15 145 90~~

~~Catalytic Naphtha 1,530 2 Reforming 470-540 108 41 105 25~~

~~Hydro- Naphtha- cracking Vacuum 260-450 40 4 10,000 3.5 70 20 Residue~~

~~Hydro- Naphtha- 260-430 20 64 1,100 10 100 190 desulfurlzation Lub Oil~~

~~Fluidized Light Oil- Catalytic Heavy Oil 465-520 27 10 20 10 130 30 Cracking~~

~~Vacuum Visbreaking Residue 450-540 58 -0.2 1,080 5 140 0 TABL I I ETemperatur e Range Head an st Quantities Require f Petrochemicao d l Processes~~

Required Energy Corresponding Thermal Output of Representative Process Feedstock Process (Per Produc ton1 t ) Plant.Capacity HTR (MWt) (Effectiveness 100%) Temp.°C Fuel Steam Electric Equivalent Steam & (10b Kcal) (ton) Power (KWH) (10* t/y) of Fuel Electric Power Ammonia Naphtha- - 780 3.6 - 14 50 265 (Steam Reforming) LNG ~800 2.5 16 185 -

~~Methanol Naphtha- ~850 4.4 - 45 30 195 2 (Steam Reforming) LNG ~850 2.1 0 95~~

~~Ethylene Naphtha- -870 6.6 70 40 390 5 (Thermal Cracking) Ethane -880 5.5 - 30 325 2 Propylene (Dehydrogenation) C3 LPG -650 1.6 3.5 50 10 25 32~~

~~Butène IsobutyriG nLP C (Dehydrogenation) 4 ~650 2.7 -2.1 -25 5 20 -10~~

~~Vinyl Chloride Monomer Ethylene ~500 1.1 1.5 120 15 25 25 (Oxychlorination)~~

~~Benzene Styrène -500 10 2 Ethylene 1.5 0.4 88 5~~

~~Mixed Xylene Separation Xylene ~550 5.4 0.2 270 10 80 5~~

~~Phenol Benzene ~300 -10 Propylene 4.5 -0.4 305 10 70~~

OJ OJ The processes in petroleum refining shown in TABLE I require comparatively mild temperatures between 250°C-550°C and heae th t processe e e froth HTGRe us n th mn I .ca s relatino t g petrochemistry showprocessee th TABLn n i , II E s requiring high temperatures (600-880°C) steae , th suc ms ha reformin g process for ammonia and methanol, and the dehydrogenation process for propylen butylènd heae an th t alsn e froeca HTGRe ous mth .

Specifically thermae ,th l cracking proces producinr sfo g ethylen procese energf th eo needd t syan lo temperatursa s i e in the range of heat obtained from the HTGR. To increase the conversion ratio, however, it is necessary to increase the heat input rat y increasineb temperature th g heatine th f o eg medium abouo ut p t 1100°C becaus reactioe eth n tim s restrictei e r fo d high selectivity.

Steam reforming also has a high process temperature level. heatine Th g medium temperatur 950-1000°f eo adequats Ci e because the restriction in reaction time of the reforming proces less si s severe than tha ethylenf o t e productioy nb thermal cracking.

~~In summary, nuclear heat utilization is highly feasible for the following processes:~~

~~- Steam reforming process - Petroleum refining process - Process for producing propylene and butylène by dehydrogenation reaction of C_, C, and LPG Proces- producinr sfo derivative, C g s~~

FIG show1 . n examplsa coaa f lo e chemistry systeme Th . nuclear proces hige s th heah n temperaturi t e userangee b dn sca for hydrogen production, ammonia synthesis, methanol synthesis uppe e s showth a figuree rn , ni th halon e anf o Th o fs d. processe lowee th w temperatur figure rn lo si th hal e f eus o f e heat.

~~332 Urea Nitric Ac Id Ammonium Nitrate~~

~~Fo rma 1 in Proces1 a s rm •••o F n i~~

~~MTBE Pr OC « S 1~~

~~Anutto n l um Suif i d e P/r id In« Plcol in«~~

~~•iCaprolactam Process~~

~~'•> S t y r e n « Process~~

~~'ri Ca r bo l i c Ae l d 011 —i Carbolic Ac id l o j o r« c ho t r O Xylenoi Naphtha l«nl Oi a Naphthalene e n 1 1 o n 1 Qu Ab sorptioa Oil~~

~~An thracene 01~~

~~i :-i Coke~~

~~FIG. 1. An example of coal chemistry system.~~

mentiones A d chemicaabovee th n ,i l industry, nuclear process heat in the high temperature ranges can effectively be used in the processes of production of hydrogen, reducing gas, ammonia, and methanol. On the other hand, the CIS process, a

333 water splitting process which uses CELOH, HI_, and H SO, in its chemical reaction cycle, whic beins hi g develope JAERy b d I (the Japan Atomic Energy Research Institute) is one of the promising processes in producing hydrogen by thermochemical water splitting.

Sinc thesl eal e processes hav esteaa m reforming a uni s a t s feli y uni ke t i tnecessar exteno t y developmene th d e th f o t helium-heated steam reformer (1), (2) , ) carries (4 a (3)t , ou d part of the National Project of the Nuclear Steelmaking (5), proceed an ) mako t d(6 demonstratioea n uni thao s t t future nuclear heat utilization can be realized.

~~2. POSSIBLE APPLICATIONS IN CHEMICAL INDUSTRIES~~

If the heat exchanger type steam reformer, which is the interface between the HTGR and chemical plant, is demonstrated ocommerciana l scale productioe ,th n processe ammoniar sfo , methanol and oxosynthesis gas will have a good chance of realization witreformere hth .

In this case, the primary helium outlet temperature from the HTGR largely affects the reforming efficiency and the nuclear heat utilization factor e factoTh . r decreases wite th h helium outlet temperature from the HTGR as shown in TABLE III. It also shows that the factor changes with the secondary helium temperature returning to the intermediate heat exchanger (IHX). A stablee showth n ,ni raisin secondare th g y helium temperature returning to the IHX, if the design condition of the HTGR allows it, is worth taking into consideration.

Nuclear heat utilization for endothermic reaction processe temperature th n si e range lower tha steae n th tha mf o t reformer processee , th suc s ha producinf so g propylend ean butylène by thermal cracking of saturated hydrocarbons, is also promising, especially when using these materials as feedstock i nseriea petrochemicaf so l plant derivativesr sfo .

~~334 TABL I HeliuEII m Temperature Nuclead san r Heat Utilization Factor~~

~~y Process HeUum Secondary Helium Gas Nuclear Heat Temperature Temperatures Temp. Utilization~~

~~HTGR EXIT IHX RETURN IHX EXIT SR IN SR EXIT m Tp CC) TjCO T5(°C) T2CO T3(°C) T4CO~~

~~1,000 290 950 930 650 860 42 950 900 880 810 38~~

~~900 850 830 760 32 850 800 780 710 25~~

~~1,000 390 950 930 650 860 50~~

~~950 900 880 810 45 900 850 830 760 39 850 800 780 710 32~~

~~1,000 490 950 930 650 860 61~~

~~950 900 880 810 56~~

~~900 850 830 760 50~~

~~850 800 780 710 42~~

~~Primary Helium Secondary Helium Process Gas~~

~~High Temperature Gas Intermediate Heat Steam Reformer Cooled Reactor Exchanger (HTGR) (IHX) (SR)~~

~~Steam Generator (SG)~~

~~e followingTh , however probleme th e ,ar s thate b nee o t d solve r thesfo d e plant configurations:~~

- A fundamental rearrangement of energy balancing is needed for introducing nuclear heat because energy-saving, effective eus of heat, or improvementof consumption rate (by combustion of off-gas or purge gas in heating furnaces) is highly achieved in present chemical plant design.

335 chemicaf I - regulatee lb o plantt e nucleas a dsar r facilities, different factors, such as periodic inspection, decrease in plant availability, safety design aseismid ,an c design will affect economy, making the present chemical plants uneconomica n desigli d operationnan .

~~3. POSSIBLE APPLICATIONS IN IRON & STEEL INDUSTRY~~

Annual production of steel when the National Project of Nuclear Steelmaking was started was 120 million tons, but the present figur hoverins ei gmillion-to0 aroun10 e th d n level. This is because the double oil shock has changed the progress of the nation's econom stablo t y e growth d furtherAn . , because th f eo chang industrian i e l backgroun d effortan d energn i s y saving, energy consumption in the iron & steel industry has remarkably decreased r thesFo .e reasons, althoug dependt i h hign o s h grade coal and iron ore, iron making by the blast furnace will play a primary role for the time being.

The most feasible way of applying nuclear heat in the future iron & steel industry will be utilization of secondary energy from a secondary energy center instead of exclusive use of nuclear heat iroe fo th rsteen& l industry thin I .s case, competitive pricf eo energy compared with fossil fuels and a stable supply of abundant energy are the keys to realization.

~~. 4 FORM SITINF O S HTG R PROCESD FO GRAN S PLANTS When using nuclear heat froHTGe procesn th mi R s plants, generally, the following two forms of siting are considered:~~

~~- Adjacent Siting - Independent Siting~~

In the former siting, the HTGR is located adjacent to the process plant to supply high temperature helium and steam directly by a piping system. In the latter siting, after nuclear heat from the HTG transformes Ri d int secondara o y energy form (e.g. hydrogen energy), it is transferred or transported to remotely located process plant used san d independnetly.

~~336 (1) ADJACENT SITING~~

~~The distinctive feature adjacenf so t s followssitina e gar :~~

~~Energ- y loss accompanied with energy conversio head nan t transmission can be minimized and economy as a whole is improved.~~

~~- Nuclear heat can be utilized in different forms of supply (process heat, high temperature steam, electric power, etc.) in response to the user's energy demand.~~

Whe- n organizin grouga complexef po s consistin differenf o g t processes utilizatioe ,th nucleaf no r heat coverin wida g e range of temperatures from high to low can be achieved reasonably and efficiently. Also becomet ,i s possibl mako t e econcisa e plant layout wit HTGe hth R locatecomplexese centee th th f n ro i d .

~~However, an investigation is necessary on the following points:~~

~~- The scheme of coupling the nuclear facilities and the chemical plants with a piping system on the same site requires strict verification of safety and reliability.~~

~~- As the HTGR is regarded as utility facilities for the chemical plants, load following operation that respond chemicao t s l plant operation is required for the HTGR.~~

~~- The problem of conformity between the nuclear facilities and the chemical plants lawarise e d regulationsth san n si .~~

~~(2) INDEPENDENT SITING~~

~~The distinctive feature isolatinn si procese th g s plant from the nuclear facilities are as follows:~~

~~- Safe decoupling can be expected.~~

~~- The consideration of energy loss during transfer or transportation of the converted secondary energy medium is not necessary.~~

~~337 Depending on the forms of the converted energy selected, a variet f combinationo y chemicaf so l plant possiblee sar .~~

In addition, provision of a secondary energy center which supplies hydrogen, reducing gas, methanol, or electric power is possible. Especially transformatioe ,th n into methanoa s la secondary energy is promising from the viewpoint of transportatio utilizationd nan .

~~5. CAPACIT HTGF PROCESR YO RFO S HEAT APPLICATIONS~~

Considering the controllability of the HTGR, the possibility of interruptio nucleae th f no r heat suppl regular fo y r inspections or shutdowns owing to causes from the chemical plant side, or the adaptability to the unplanned increase of production capacity in the future, the modular scheme, with an HTGR of appropriate scale, superios i n installatioa o t r n wit largha e HTGR.

In this case, the most appropriate capacity for one HTGR determinemodule b n ca e d froHTGR'e th mn safet sow y features sa operationae welth s la l factors productioe , th suc s ha n capacity of the chemical plants.

HTGe Th recognizes Ri havins a d g many safety advantagest ,bu for use in the chemical industries, intrinsic safety and passive safety without active components will strongl requiree yb d from the safety point-of-view e capacit HTGTh e .on Rf modulo y s i e determined from the viewpoint of effectiveness in passive safety.

~~6. COMMENTS ON ECONOMIC EVALUATION OF NUCLEAR PROCESS HEAT APPLICATIONS~~

~~Regarding nuclear substituta hea s a t hear fo et froma heating furnace in the chemical industries, an economic comparison by usin cose unir th gpe t t hea mads i ts follows ea :~~

~~- Conventional method (Use of fossil fuel) Fixed cos heatinf o t g furnac Fuee+ l cos Othe+ t r operation cost~~

~~338 - Nuclear heat utilization method Fixed cos HTGf primard o t Ran y heat exchange HTGr+ R operation cos Fixet+ d cos procesf o t s heater Othes+ r operation cost~~

assumptioe Baseth n o d n thafixee th t d cos procesf o t s heaters using nuclear hea fixee equas th i t e do t lth cos f o t heating furnace economin ,a c compariso mads ni e betwee cose nth t per unit heat quantity of fossil fuel and the fixed cost of the HTGR and the primary heat exchanger including the operation cost HTGRe oth f .

TABLE IV shows a forecast of the fossil fuel costs. For the HTGR to be competitive, the (fixed cost of HTGR and primary heat exchanger + HTGR operation cost) per unit heat quantity must be less than the value of the fossil fuel shown in TABLE IV.

~~TABL V ForecasEI Fossilf to e Fuel Cost wit Escalatio% h7 n~~

~~(Yen/1000 Kcal) ^"\^^ Year Classifï5»^^^ 1984 2000 2020 2040 2060 cation ^^-^^^~~

~~Imported Coal 1.8 5.3 20.6 79.6 308~~

~~Naphtha 5.4 15.9 61.7 239 924~~

~~Kerosene 7.7 22.7 88.0 340 1,320~~

~~Heavy Oil (A) 7.0 20.7 80.0 309 1,200~~

~~LPG 6.7 19.8 76.5 296 1,150~~

~~L N G 7.5 22.1 85.7 332 1,280~~

~~Electric Power 20.2 59.6 230.8 893 3,460~~

~~339 7. CONCLUSION~~

~~This paper present srecena tperspectiv e studth n yo f eo HTGRs in the chemical and iron & steel industries of Japan. As for future applications, the following are concluded:~~

- A secondary energy center which supplies secondary energy such as hydrogen, reducing gas, methanol electrid ,an c powes ri feasible. Especially transformatioe ,th n into methanoa s la secondary energ promisins i y g froviewpointe th m f so transportatio d futurnan e utilization.

Whe- I chemistrnG aromatid yan c chemistry basia ,n whici c e har study presene stagth t ea t time, achiev prospece eth f o t commercialization, nuclear heat utilization could e extenth o t d whole petrochemical industry and the concept of a petrochemical center will be feasible.

- The method of producing hydrogen by decomposing water with the nuclear heat, will become promisin neae th r n futurei gt i d ,an is considered that the chemical industry will develop a new configuration using hydrogen energy which will be a most important innovation. mose Th t - feasibl applyinf o y ewa g nuclea iroe r th hean& n i t steel industr utilizatios i y secondarf no y energy froma secondary energy center. In this case, the competitive price of this energy compared with the conventional method and the stable suppl abundanf yo t key e energrealizationo th st e yar .

~~REFERENCES~~

~~(1) T. MIYASUGI, S. KOSAKA, K. MUKAI and A. SUZUKI, "Characteristic Heat-Exchangea f so r Type Steam Reformer,. "J Japan Petrol. Inst., 25. No. 4 (1982), 260~~

~~(2) T. MIYASUGI, S. YOSHIOKA, S. KOSAKA and A. SUZUKI, "Experimental Study of a Heat-Exchanger Type Steam Reformer with a Low Steam/Carbon Ratio. Effect of Carbon Deposition~~

~~340 on the Distribution of the Process Gas and the Temperature Distribution in the Catalyst Tubes," Inst. Chemical Engineering J26, No. 1 (1986), 130.~~

~~(3) T. MIYASUGI, S. KOSAKA and A. SUZUKI, "Dynamic Analysis and Simulatio Helium-Heatea f no d Steam Reformer, Canadiae "Th n Journa Chemicaf lo l Engineering (1982)0 ,j> , 759.~~

~~YASUKAW. S MIYASUG. T ) d (4 Aan al. t Ie , "Economic Evaluation of Helium-Heated Steam Reformer", Presented at the Integrated Energy System Meeting held in Tällbeng, Sweden, 1-5 September, 1986.~~

~~TSURUOKA. K ) (5 INATANI. ,T MIYASUG. , T MIZUNO . M d Ian , "Design Study of Nuclear Steelmaking System", Transactions of the Irod Steenan l Institut Japanf eo , Vol , 1983.23 , PP1091-1101.~~

~~. INATAN T TSURUOK. K d ) Ian (6 al.t Ae , "Process Heat Application System of High Temperature Reactor", Alternative Energy Sources IV, Vol. 6, 1981 PP15-29.~~

~~341 ISRAELI PERSPECTIV HTGRN EO s~~

~~A. BARAK, A. BECK, E. GREENSPAN, J. SZABO Atomic Energy Commission, Tel-Aviv L. BLUMENAU, H. BRANOVER, A. EL-BOHER, . SPEROE . SUKORIANSKS , Y Ben-Gurion Universit Negeve th f yo , Beer-Sheva Israel~~

~~Abstract~~

A preliminary assessmen interese th f to t Israel should HTG e havth n ei R technology, and of possibilities for improving the efficiency of HTGRs using the Liquid Meta(LMMHDD lMH ) energy conversion technolog undertakens yi . Modular HTGRs were foun offeo dt collectioa r featuref no s which make them attractiv Israele th n ei i conditions. These features include inherent safety, resistance against acts of war, investment protection, suitability to the Israeli seismic conditioncoolingy dr o t d , possibilitsan optimar yfo l developmene th f to Israeli grid and, hopefully, for economical sea-water desalination and oil-shale utilization. Being fre f rotatino e g machiner beind yan g applicable ove entire th r e temperature range of HTGR energy delivery, the LMMHD technology opens promisinw ne g possibilitie utilizinr sfo hige gth h temperature operatio nabilit- y of HTGRs for the production of electricity. A number of novel LMMHD cycles were conceive attractivd dan e matche unique th o t se HTGR heat source characteristics were identified; they enable producing electricity with efficiencies whic ideae th h f lexcee o efficiencie % d70 sver e eveth yn i hig h temperature domain r exampleFo . LMMHe th , D technology might enable attainint gne respectively, in % 47 d efficienciefirssecone d an ,th an t % d 43 generatio f so f no HTGR s comparea s d attainablwit% 40 h o efficienciet e % wit37 hf o s conventional steam energy conversion technology.

343 . 1 INTRODUCTION Of the several advanced nuclear reactor concepts emerging in recent years, modular High Temperature Gas Cooled Reactors (HTGRs) attracted particular attention in Israel. This is due to a collection of a number of interesting features, includin followinge gth : inherent safety, small unit size, adaptabilit cooliny dr potentiad o yt gan varieta r lfo industriaf yo l applications in addition to the production of electricity. Consequently e Israeth , l Atomic Energy Commission (IAEC) undertoo kpreliminara y assessmen interese th f o t t Israel should e havth n ei HTGR technology. In parallel, the IAEC encouraged the Center for Magneto-Hydro-Dynamic (MHD) Studies (CMHDS Ben-Gurioe th f o ) n University (BGU investigato )t e possibilitie improvinr sfo efficience gth f yo HTGR productioe th r sfo electricityf no , capitalizin hige th h n gtemperaturo e operation abilit bot f Liquiye HTGo e th h th d dRan Meta (LMMHDD MH l ) energy conversion technologies. The present paper reports upon the preliminary findings obtained so far. It consists of two distinct parts: the first part (Sec. 2) refers to the IAEC perspectiv HTGRn eo s wherea secone sth d part (Sec ) report.3 s upoe nth preliminary findings from the study of the promise of the LMMHD energy conversion technology for HTGRs.

2. THE INTEREST ISRAEL MIGHT HAVE IN MODULAR HTGRs 1 2. Implication Developmene th n so Israele th f to i Electricity Generation System The expansion of the electricity generation system in Israel is more problematic than in many other countries due to its (1) Relatively small size - maximue th m peak deman arouns di d 2600 Isolatio) MWer (2 fa d o s an , n- electricity bu t no Israe yn froca l s neighboursmit a suddecasn i ,f eo n generation shortage. These constraints impose penaltreserve th n yo e

344 generating capacity that need be installed in the country and, hence, on the cost of electricity in Israel. Modular HTGRs (MHTGR) might enable expandin Israele gth i grid with minimum over-capacity. This is due to the following reasons: (1) Minimum overshooting of the total reserve capacity needed. Consider, for example, the two options that were recently considered for further expansion of the ~4000 MWe Israeli grid: adding 550 MWe coal LWRe smallese fireth MW s- df prove 0 o unitt95 R sizr nso eLW desig n commercially available. As the average increase in the peak demand of electricity in the mid-nineties is expected to be ~150 MWe per year, the average overcapacity (beyond the reserve capacity) can be of the order of 200 MWe and 400 MWe for, respectively, the coal and nuclear power plants. If the total installed capacity in the mid-nineties will be ~5000 MWe, an extra corresponde MW averag0 n a 40 o st e capital cost ~8% penalto t p u . f yo (2) Reducing the installed reserve capacity needed. The installed reserve capacity - aimed at replacing long-term outages, depends on the largest size unit installed in the grid. Thus, if the 950 MWe LWRs were to be installed in Israel, the installed reserve capacity would have othee leas t th MWea 0 rn e o t95 handb , o If t . , MHTGRs wil installede lb e ,th reserve capacity need be only ~550 MWe - the size of the largest coal plant in the system. The extra 400 MWe required for the reserve corresponds to an additional capital cost penalty of up to ~8%. (3) Reducin non-spinnine gth g reserve capacity. This reserve capacit tako t s eyi car shorf eo t outage ordee a th f f ro s(o fraction of an hour) - it consists of inexpensive units such as gas turbines. The smaller the largest generating unit size, the smaller need be the non-spinning reserve. (4) Reducin spinnine gth g reserve. This reserve is to take care of outages for very short periods (of the order of a few minutes) - it is obtained by operating a fraction of the units at

345 partial load. Partial load operation usually involves cost penalty due to reduced efficiency. The smaller the largest generating unit size, the smaller this penalty is expected to be. (5) Improving the grid's load following capability. The need for a spinning reserve forces at least a fraction of the generating units to operate at least a fraction of the time at below rated capacity theio t e r relativelDu . operatiow ylo n costs, nuclear o t unit e ar s operate at the nominal capacity. Hence, the larger the fraction of the installed nuclear generating capacity in the grid, the larger becomes the load reduction of the fossil fuel fired power plants. Since the electricity generation efficiency decreases faster than the load reduction, the smaller the overcapacity nucleaf o r generating unit more sth e efficient averagee th n electricito e , th , y generatio expectes ni . be do t Additional economic attributes of MHTGRs include the following: ) (6 Short installation period. Obviously, the shorter the installation period the lower is the interest during construction. (7) Reduced uncertainty in commissioning time. Standardization, factory fabrication and shorter fabrication and construction time expected for the MHTGRs reduces the probability that the power unit wil untimele b l y commissione deithe- r becaus inaccuratn a f eo e deman ddelayo t forecas e installationn si du r o t . Late commissioningy ma impos penaltea fore th shortagf m o n y i electricitn ei y supply earlo To y.a commissioning will impose a capital cost penalty. (8) Increased availability. Generally speaking highee th , availabilite rth smallee yth re neeth e db installed capacity. On-line refuelling, low radiation levels, system simplicity, standardizatio d smalan n l size component e expectear s o makt d e th e availabilit HTGRf yo s significantly higher tha conventionaf no l LWRs.

346 Take l togethernal appeart i , s that MHTGRs might enable developing the Israeli grid in an optimal way. The cumulative economic benefit resulting fro above mth e outlined attribute vere b yn significantsca . Whethe thit no s r ro benefit will make MHTGRs economically competitive with other types of power sources in Israel is an open question, as yet.

2.2 Safety implications and investment protection The inherent safety aspect of MHTGRs was borne out by a number of safety studies [1-5] which considered "internally" initiated accidents (i.ee du . to the malfunction or failure of components, or to operator errors) as well as certaio t n "externally" initiated accidents, suc loss h a f offsit so e powed ran earthquakes. As part of the assessment of the interest Israel should have in HTGRs, we set to examine the implications, on the safety of modular HTGR's, of yet another typ f "externallyeo " initiated accidents namely, accidents resulting from acts of sabotage or acts of war. Also examined are ways to reduce the damage by such acts and their consequences. The assessment is done with reference to the MHTGR power plant proposed in Ref. 6, in which the NSSS is located underground, inside a silo. preliminare Th y assessment carried-ou standare th n o t d underground design, modified to have some additional, relatively inexpensive sheltering features, coul t identifdno credibly f sabotagyan o whicr t eac wa r heo could bring about excessively high (i.e. >1600°C) fuel temperature and, hencea , significant release of fission products. Turning, next, to the issue of investment risk, it is clear that having a relatively small capacity per module, the modular design is significantly more investment risk resistant than an integrated design. Moreover, a preliminary assessment indicates that by relatively simple and inexpensive design modification t shoulsi possible db mako et MHTGR'e eth s NSSS practically undamageable by any conceivable act of sabotage or act of war. The above

~~347 features are of great significance for the acceptance of nuclear power reactors in Israel.~~

2.3 Siting implications The present population density precludes the possibility for siting conventional LWRs in the Mediterranean coastal zone in Israel. However, the combination of inherent safety, relatively small unit size and resistance to acts of sabotag enably ema e siting MHTGR near Mediterraneao e y rth s b n coast. If coastal sites will be accepted, MHTGRs may have a significant economic benefit (See Sec. 2.4) on conventional nuclear power plants, such as LWRs, which may not be accepted for siting in the relatively densely populated coastal zone. Preliminary seismological studies performed on a number of possible sites in Israel indicate that nuclear power reactors will have to be designed to withstand a relatively high g-factor. Recently it was proposed that the California seismic standards be adopted for Israel. This is due to the many similarities between thes regionso e Californietw Th ) (1 : a seismologs yi Andreadominaten Sa e th s y fauldb t whil Israen ei e Deae th rif a th ls i dtSe dominant factor. (2) The San Andreas fault runs about 100 km inland parallel Pacifie th o t c shoreline throughout California whil Deae rifa eth dtSe runs along Israel parallel to the Mediterranean shoreline. (3) Many minor but potentially hazardous faults branch out from both the San Andreas and the Dead Sea rift, and (4) In both cases there is a significant amount of uncertainty relate earthquake th o dt e frequenc damage th d yean potential associated with each minor fault line. It has been reported [6] that MHTGRs designed to have their NSSS properly supported inside a silo located underground can meet the California seismic standards (of 0.5 g horizontal acceleration) with, essentially, no capital cost penalty. Standard LWRs, on the other hand, are designed to withstan dhorizontaa l g-facto f 0.1o r 0.255o t . Upgrading their desigr nfo

~~348 withstanding a horizontal ground acceleration from 0.25 g to 0.5 g is expected capita increasR o t ] [7 LW . lmuce s cosa e6% th y s htb a~~

2.4 Adaptability to Dry-Cooling Due to the relatively dense population along Israel's Mediterranean coast, coastal and near coastal sites are not likely to be available for nuclear power plants in Israel. The other major water bodies in Israel are located alon e Dea nona d riftgth an f dSe the, o e n offe mca r acceptable sites. Consequently, unles majosa r water transport system propose e (sucth s ha d canal to connect the Dead Sea with the Mediterranean) is constructed, the only practical way to discharge the heat from nuclear power stations in Israel appears to be via cooling towers. Dry, wet and dry-wet cooling towers are being considered in Israel. However, with the increasing shortage of water Israel is facing, it might be in nationae th l interes sparo t watee e th applicationr rfo whicr sfo h thero n s ei substitut r waterefo . Consequently vera , y preliminary assessmene th f o t adaptability coolingy dr o ,t HTGRf ,o . LWRsvs s unde Israele rth i conditions was undertaken. The adaptability of HTGRs to wet and dry-wet cooling towers wil examinee lb futuree th n di . Relative to direct, once-through cooling, dry cooling penalizes the economics of power plants in Israel in a couple of ways: (1) By reducing the attainable efficiency - as a result of increasing the steam condensation temperature increasiny B ) (2 capitae d g th operatin an d , an l g a cost s a s- consequence of the capital investment in the cooling towers and of the power consumption for air circulation.* Table I summarizes our preliminary estimates of the penalty thus imposed on LWRs and two generations of HTGRs. The two generations are characterized by a helium inlet/outlet

* Natural drought dry cooling towers appears to be impractical under Israeli conditions.

~~349 TABLE I. Preliminary Estimates of the Effect of Dry Cooling on the Efficienc Economicd yan Nucleaf so r Power Reactor Israeln si .~~

~~HTGR Characteristic LWR 250/700'C 550/950'C~~

Net Efficiency (%) Coastal site coolina ;se g 32 40 47 Inland; dry cooling 31.1 39.2 46.5 Relative efficiency loss 2.9% 1.9% 1.1% Installed capital cost increase ($/KWe) Due to efficiency loss 60 36 18 Cooling towers + energy needs 400 282 212 Total 460 318 230 Installed cost relative increase 23.0% 15.9% 11.5% Cos electricitf to y increas) e(% efficienco t e Du y loss 2.9 1.9 1.1 Due to installed cost increase 16.9 9.6 6.5 COE relative increase 19.8% 11.5% 7.6%

temperatur , respectivelyeof , 250'C/700°C 550°C/950°Cf o d an . lattee Th .r uses the binary cycle consisting of a wet-vapor topping cycle and a LMMHD Rankine steam bottoming cycle describe Secdn i . 3.3.1. In evaluating the implications of dry cooling it was assumed that the average effective condensation temperature is 40°C, versus 33°C with direct sea cooling; that the extra investment in dry cooling for LWRs is $400 per installed kWe (including the effective capitalized cost of the energy required for forcing air through the towers); that the total installed cost of LWR and HTGR in a coastal site in Israel is 2000 $/kWe, and that the cost of the dry cooling towers scales linearly with their capacity. The preliminary assessment indicates that the economic penalty to HTGRs associateo t d wit % usee 60 h th coolinIsrael y n e ,i b dr f n o ,g ca penalte th f almosLWRso o y t % 40 t . Moreover secone th f i , d generation HTGRs will be used for process heat applications (say for synthetic fuel

350 production (see Sec. 2.5 addition )i electricito nt y generation, their cost penalty is expected to be significantly less than 40% that of LWRs. This is due to the very large fraction (approaching, possibly, 90%) of this HTGR energy utilizee whicb n usefur hca d fo l purposes. If either gas-turbine technolog r LMMHo y D energy conversion technology (see r example,fo , Sec. 3.4.3 e developeb ) o becomdt n a e attractive match to HTGRs, dry-cooling will impose very little penalty, if at efficience th alln o , y attainable from HTGRs relativele th o t . e yThidu s si high temperature for heat-rejection from the corresponding gas cycles.

2.5 Israeli oil-shale utilization The main organic fuel resources of significant economic potential discovered so far in Israel are oil-shales. The presently proven reserves are 10 billion tons. The organic material contents in the Israeli shales is in the vicinity of 15% - significantly lower than in rich deposits (found, for example, in Colorado), but comparable to the quality of shales found in many parts of the world (such as Australia, Morocco, China and the Eastern U.S.A.) e totaTh l. amounl whicoi f o ht coul e extractedb d , using conventional methods, fro 10e mth 1 0 ton Israelf so i shalemillio0 60 s si n tons. presene Ath t t leve energf lo y consumptio countrye th n i , this correspondo st a potential for supplying the total national energy needs for a period of approximately 75 years. Conventional method extractinr sfo g liqui gaseoud dan s fuel materials from oil-shales call for burning part of the organic material of the shale in order to supply the heat required for the process. Rather than burning organic material it is possible to provide the high temperature (i.e., >800°C) heat required for the production of fuel from shales from HTGRs. A number of studies of the feasibility of so doing were performed both in the U.S.A. [8-10] and in Germany.[11] A general conclusion of these studies is that with HTGRs as the primary heat source, it might be possible to increase the

351 amount of fuel extractable from a given quantity of oil-shales by as much as ~l/3, and to do so at a comparable (in certain cases even somewhat lower) coswitd tan h significantly lower adverse environmental effects than wite hth conventional approaches. prospecte Th r increasinsfo liquie ggaseouth d dan s fuel extractable fro limitee mth d resource f oil-shalef courseso o , is 3 ,Israe n 1/ sveri y yb l appealing cruciaA . l questio s whetheni r this coul Israele donee db th n i ,i conditions witr , o wit , h hrelativelno y small economic penalty. This question will be addressed in the future. largese exploitee Th firsb d o an t t l shaldoi e Rotefiele th d- m fields i , planey northere locatedr th e f so th nn di Negev district. Sitin HTGRf go n si these planes appears to be quite attractive - from the point of view of distance from population centers quitd an , e acceptabl e- fro pointe mth f vie o f wo seismic (relative high g-factors dry-coolind an ) g design requirementss A . discussed in Sees. 2.3 and 2.4 the economic penalty these design requirements impose on HTGRs is relatively small.

6 Adaptabilit2. o Sea-Watet y r Desalination Israel suffers today fro msevera e shortag waterf eo average th ; e annual consumption exceeds the natural annual supply by a few hundred m.cu.m. (millions of cubic meters). This water overdraft has accumulated to about 1500 m.cu.m. - the consumption equivalent of about 10 months, causing overpumping from the aquifers and the Sea of Galilee. This overpumping brough e wateth t r tabl metersw efe a dangerou dowa o t ,y b n s level. Desalinatio majoe th f sea-watenf o r o possible on s ri e solution watee th o rst shortage problems. The two seawater desalination processes that seem most promising to the professional community in Israel are the Multi-Effect Distillation (MED) combined with power generation and the Reverse Osmosis (RO).

352 2.6.1 Multi-Effect Distillation procesD Th s beeeME ha s n develope Israee th y ldb Desalination Engineering (I.D.E.) compan las e year9 th 1 t n yi s drivin[12]e Th . g forcr efo this evaporation process is low grade heat; the upper practical temperature for heat suppl dictates i , yabou- °C materiay 5 db t7 l compatability considerations. The most economical source for such heat in Israel is the exhaust steam from power-plants. A preliminary assessmen promise th f o tf HTGReo r sea-watefo s r desalination reveale numbeda potentiaf ro l advantages over LWR heae th ts sa source for MED: ( 1 ) Direct coupling to the MED plant pressurAe sth primar e th n ereactorr i y(o ) coolins i gR systePW a mf o higher than the pressure in the secondary cooling (or turbine) system, any leak in the steam generators is bound to cause water from the reactor system to leak into the turbine system. Since the reactor system water may be radioactive a direct couplin e turbinth f o ge system stea planD ms i ME t wit e th h unacceptable. Instead heae th , t fro turbine mth e cycl transferres ei e th o dt desalination plant via a pressurized sea-water loop which interfaces with the turbine loop throug conventionaha l condenser. A different situation exists in HTGRs. As the primary helium system pressur s significantlei y lower tha secondare nth y coolant (i.e., steam) pressure steae lea,a th mn ki generator likelt no contaminats o ysi t turbine eth e steam. Consequently, it might be possible to directly couple the turbine exhaust steam with the MED plant. Being free of the condenser, huge circulation pump and flashing unit, such a direct interfacing is expected to be simple cheaped ran casre LWRsr thasavth o t e pumpinfo e n ni d eth an , g energy.

~~(2) Higher efficiency. The higher the driving steam temperature, the lower is the energy consumption and cost of desalination in MED plants. Conventional turbines~~

353 can operate with a backpressure as high as 5.5" Hg, corresponding to a steam exit temperature of 58.5°C. This is the steam temperature HTGRs can e poweth s i r sourceR plantLW D f ,I ME 5.5.provid e g "th H o t e condensation pressur equivalents e i indirece th o t D e t couplindu , ME e th o gt plant, to a ~54°C driving steam, resulting in a reduction of almost 20% in the water production per unit of the turbine exhaust steam. (3) Using gas-turbine with high heat rejection temperature. HTGR attaiy additionasma n n a l significant efficiencf o e us ye gaith y nb turbines ga s instea steaf do m turbines. HTGR-GT reject their unused heaa t a t relatively high temperature. This reject heat could provide ~75°C steae th mo t MED plant with no or very small penalty on the electricity generation efficiency °C5 7 steae producn Th m.ca e twic mucs ea h desalinated wates ra produceabl same th ey e b steamquantit R LW .f yo (4) Higher availability. economice Th desalinatiof so n favors high availabilit higd yan h capacity factor. As discussed in Sec. 2.1, MHTGRs are expected to have higher capacity factors than LWRs. (5) Coastal siting As discusse Secdn i . 2.3 mucs i t ,i h more likely that MHTGRs coule db sited by the Mediterranean coast than LWRs. Coastal or near coastal location has a significant economic attribute for sea-water desalination.

2.6.2 Reverse Osmosis The reverse-osmosis desalination process is energized entirely by electricity producn ca m.cu.mt 0 I MWe.0 e13 10 year r .pe . rpe Thu 0 s70 providn ca presene e th ef o abou MW t wate% t50 r consumptio Israelf no s A . the RO units are suitable for load shedding, the electricity production capacity destined for driving the RO plants could serve as the spinning as well as non-spinning reserve capacity for the Israeli electricity grid, and possibly even a installese parth f to d reserve capacity (see Sec. 2.1). Suc multi-purposha e

354 schem likels ei improvo yt overale eth l economic electricitf so y generatiod nan water desalination in Israel. Whether or not this scheme could make water desalination economica Israe n opeli n a s nli question yets ,a .

7 Conclusion2. s The preliminary assessment indicates that modular HTGRs offer a collection of features which might be very attractive for the Israeli conditions. Among the potential attributes of MHTGRs identified are the following: ( 1 ) Very high level of public safety. (2) Resistance against acts of war and sabotage with negligible economic penalty. (3) Better investment protection ability than conventional nuclear reactors. (4) Possibility of meeting the demand for electricity with minimum overcapacit minimud yan m reserve capacity. (5) Better suitability than conventional LWR seismie th o st c conditionn si Israel. (6) Better suitability than LWRs to dry-cooling in Israel. (7) Higher likelihoo coastar dfo l siting than conventional nuclear reactors. (8) Higher efficiency than conventional LWRs as an energy source for sea-water desalination. (9) Potential for significant improvement in the utilization of the Israeli oil-shale resources.

All combined, these attributes might significantly improve the economics of MHTGRs in Israel. Whether or not this improvement will make MHTGRs economically competitive with other power source , howeversis , not clea yets ra . Wha cleas ti thas inherenri e tth t safety featur MHTGRsf eo , combined with their resistanc acto e t sabotag f so ward ean , small unit sizd ean adaptability to siting in the sparcely populated parts of the arid Negev district

~~355 giv MHTGn ea R type reacto bese rth t chanc beinf eo g accepte Israee dn i th n li post Chernobyl era.~~

. 3 POSSIBILITIE IMPLICATIOND SAN LIQUIF SO D METAD LMH ENERGY CONVERSION IN HTGRs 3.1 Introduction Unique feature s theif HTGRsa o s r r fa energ s a , y conversios i n concerned, is that (a) their helium coolant can be provided at very high temperatures - possibly exceeding 950°C, and (b) they deliver their energy ove relativelra y very large temperature 650° AYRo range t th p u Cn ei - . [13] The latter feature is a consequence of two constraints: (1) Relatively low volumetric heat capacit gaseoua f yo s coolant and particulan ,i Germae th n ri n HTGR design approach,[13] (2) The need to maintain, with the primary coolant metae ,th l structural component reactoe th f sbelot o ra w ~300°C. HTGRs destined for power production are usually designed to have a conventional superheated steam Rankine cycl r theiefo r Power Conversion System (PCS), yielding a net efficiency approaching 40%.[13] The cycle upper steam temperatur f ~550°Ceo materialy b t se , s strength limitationss i , well below the helium outlet temperature attainable. Thus, the conventional Rankine cycle does not make a good thermodynamic match to HTGRs. Moreover probabilite th , water yfo r ingress fro secondare mth y water system int primare oth y syste safeta ms i y hazard.

Gas Turbines (GT) were proposed [13] for utilizing the above 550°C operation abilit HTGRf yo power sfo r production predictee Th . d efficiencf yo HTGR-GT was 40%-41%, pertaining to a helium coolant outlet temperature of 850°C imposed by material strength limitations. A significantly higher efficiency of -47% has been calculated for HTGR-GT operating with a binary cycle consisting of a gas-turbine (i.e., Brayton) topping cycle and an organic coolant Rankine bottoming cycle. [13]

356 Despite the inherent simplicity of the direct-cycle HTGR-GT and to the high efficiency expected from HTGR-GT havin a binarg y cyclee th , developmen f HTGR-Go t bees Tha n abandoned e bot th Europn hi n i d ean U.S.Atechnologicalo t e du . safetd an y difficulties encountered, including: [13] turbomachine integrity; lubrication oil ingress; seal design; primary system acoustic level; materials and maintenance considerations. All of these difficulties are associated with the rotating nature of the GT technology. Being free of rotating machinery and being applicable over the entire temperature rang HTGf eo R energy delivery LMMHe th , D technology opens possibilitiew ne utilizinr sfo hige gth h temperature operation-abilit HTGRf yo s productioe foth r electricityf no purpose Th . thif eo s sectio reporo t s ni t upon preliminary results fro e searcmth r possibilitiefo h f improvino s e th g efficienc HTGRf yo s usin LMMHe gth D energy conversion technology. Starting wit hbriea f descriptio principlee th f no f LMMHso D energy conversion (Sec. 3.2) we describe a number of novel LMMHD components and cycles (Sec. 3.3). Section 3.4 describes a number of possibilities for matching PCS, based on these and other LMMHD cycles, with HTGRs and estimates the attainable efficiencies. Other implications of coupling the LMMHD energy conversion technology with HTGRs are discussed in Sec. 3.5.

3.2 Principles of LMMHD energy conversion principle Th f LMMHeo D energy conversio illustrates ni Fign di 1 . which shows an MHD generator of DC-Faraday type. This generator consists of a channel through which an electrically conducting fluid flows perpendicular to a magnetic field, creating an induced electric field. By connectin electrodee gth loada o st close,a d pat creates hi d through which current can flow. The power produced by the MHD generator is the product of the load voltage and the current. As gases have very low conductivities, they must be heated to about 2000°K and seeded with materials having low ionization energie achievo st e adequatD e MH conductivitie n a n i e us r sfo

~~357 Magnetic Fi «Id Load~~

~~Channel~~

~~Current Liquid Metol orTwo x Phase Flow \~~

~~generatoD FaradaC MH D Fig f r1 o .y type~~

generator exampler Fo . , helium seeded with 0.45 atocesiuf o m% m will have conductivita roughlf yo y lO(Qm)" 2000°t 1a atmcontrast 1 y B d . Kan , liquid metals have conductivities of the order of 106 - 107 (Qm)"1 at low temperature. This implies that LMMHD generator significantle sus y lower flux density and cost less than their plasma MHD counterparts. Moreover, LMMHD generator e driveb HTGRn y b nca s s whereas plasmD MH a generators cannot.

3 3. Novel LMMHD cycle componentd san s A description of the LMMHD approach to energy conversion and a review of the evolution of this technology since its inception in the late 1950s, ca foune nb Ref. n dReferenci 14 . provide5 e1 reviesa novewf o l LMMHD components and cycles recently conceived at the BGU CMHDS. The following is a brief description of selected cycles of relevance to HTGRs.

3.3.1 Wet-Vapor Cycle A simplified LMMHD cycle operating in the "wet-vapor" regime is illustrated in Fig. 2. The working fluid can be any liquid metal having proper vapor pressure at the temperature range of interest. Heat is transferred, in an isobaric process, from the source to the working fluid bringing it to its

~~358 Two Phase MHD (0) Generator (b) 2~~

~~Heat Source~~

~~£-, MHD Pump~~

~~NetPower(DC)~~

~~Fig. 2 A schematic layout of a single stage wet-vapor homogenous LMMHD cycle (a) and the associated T-S diagram (b)~~

saturation point (Process 4-1, Fig. 2). The fluid is partially flashed (i.e., a small fraction of it boils) isenthalpically before entering the two-phase MHD generator (point 1). As the gas-liquid mixture expands isentropically through the divergent MHD duct (1-2), work is extracted. As the pressure drops, more liquid vaporize temperature th d san e falls. After expansionw lo e th , temperature low pressure saturated fluid is cooled (2-3) and the vapor condenses liquie Th pumpes .di d back systee (3-4th o t )m high pressure before entering the heat source. Unique features of the wet-vapor cycle include the following: (1) It can use a single working fluid; its liquid phase serves as the electrodynamic (i.e. high conductivity) fluid whereas its vapor phase serves as the thermodynamic fluid. In contrast, all other LMMHD cycles propose differene dus t material r theisfo r electrodynamid can thermodynamic fluids. (2) It is free of mixers and separators and, hence, of the kinetic energy losses associated with the operation of these components. neee regenerativ r freth s i d fo f t eo I ) (3 e heat exchangers. (4) It can make a perfect match to heat sources which deliver their energy over the entire temperature range of the cycle.

359 3.3.2 LMMHD compressors and Ericsson cycles By operating a two-phase LMMHD generator in the reverse (i.e., analogously to the operation of a LM electromagnetic pump), it is possible to vapor compreso s rga beine sth direcn gi tresultin e contacTh . tg witLM e hth compressor has two unique properties: (1) It performs the compression nearly isothermally. (2) It is free of rotating or reciprocal machinery and is free of the need for sealing shafts (i.e., it can be hermetically sealed and nearly maintenance free). LMMHe Th D compressor, when combined wit LMMHha D generator, enables the design of gas cycles characterized by nearly isothermal expansion and compression. Fig show.3 schematisa c layouresultine th f o t g Ericsson

~~LM HTR~~

~~M MHO Pump '(I-C2' LM~~

~~Mixer~~

~~Separator~~

~~NOTE=With Gas Heating 83"I i With Liquid Metal HeatingE,*!~~

~~Fig 3 . Ericsson cycle schematic witdiagraS hT- m~~

360 cycle with the corresponding T-S diagram of the thermodynamic working fluid (i.e. gas, in this cycle). The isothermal expansion and compression mak theoreticae eth l efficienc thif yo s cycle Carnoclose th o et t efficiencyn I . contrast, conventiona cycles lga s base turbomachinern do Braytoe th f o ne yar type; the characterizee yar adiabatiy db c expansio compressiod nan e ar d nan less efficient than Ericsson cycles operating betwee same nth e temperature limits. Ericsson designecyclee b n sca havo dt variablea e fractio heae th tf no source energy delivered to the electrodynamic (i.e., LM) and thermodynamic (i.e., gas) fluids. In the following we shall assume that all the HTGR power

~~is transferred to the gas, i.e., that E2=l. After being heated by the HTGR~~

(point 9), the fraction e{ of the PCS helium is mixed with the other fraction (l-£j) whic regenerative hth cornef o t sou e heat-exchangee r (poinTh . 8) t combined high-temperature high pressure heliu mmixes i o t dM L wit e hth provid two-phasea e mixture wit propeha r inlet void fraction. Nearly isothermal expansion through the MHD generator channel to the maximum void fraction follows (1-2), with subsequent separation of the liquid and gas. The low pressure gas is then cooled in the regenerators in two steps: first agains hig e fractioe th h tth f pressuro j n£ (2-3s thed ega n)an agains entire tth e hige floth h f wo pressur temperaturw elo s (3-4) reverse ega e Th th . f eo expansion process, isothermal compression, occurs in the LMMHD compressor (5-6). The two phases are again separated and the high pressure gas return hige th h o st temperatur e mixer (poin afte) 10 t r being heatee th n di

regenerative heat exchangers and in the HTGR. By varying the fraction el of the helium that goes through the HTGR, it is possible to adjust the helium temperature difference acros reactoe consistene sth b o rt t wit reactoe hth r design requirements.

~~361 1000~~

~~(a) (b) 900~~

~~800~~

~~700 ? ~ta 600~~

~~D o 500~~

~~E 400 Q> l—~~

~~300~~

~~200~~

~~100~~

~~0 0 1 S O 6 0. 4 0. 2 0. 0 8 0. 6 0. 4 0. 2 0. 0~~

~~Fraction of Heat Delivered~~

Fig 4 Temp.-are. a diagra 300-950'a f mo C HTGR drivin reheae on ga t conventional Rankine steam cycle (a), and a binary cycle (b) consistin homogenoua f go s toppin supercriticaa gd cyclan ) e(T l steam Rankine bottoming cycl) e(B

3.4 LMMHD PCS for HTGR 3.4.1 Mercury Wet-Vapor Topping + Rankine Steam Bottoming PCS Conside HTGn a r heliue Rth m inlet/outlet temperatur f whiceo s hi 30(rc/950°C. Fig. 4a shows the temperature dependence of the heat delivered fro primare mth y (helium secondare th o )t y (steam) coolants when conventional Rankine steam cycle having one internal reheat stage is used for the PCS shadee Th . d area represents exergy (and, hence, efficiency) loss. One way to reduce the exergy loss is to interpose an LMMHD wet-vapor cycl between ei heliue nth m hea tRankin e sourcth d ean e steam cycle. The temperature-heat diagram of the resulting binary cycle is shown in Fig. 4b. Mercury is assumed to be the working fluid of the topping cycle due vers it yo t favorable vapor pressur temperature th n ei e range consideredo T . limit the number of secondary coolants, mercury exchanges heat with the

362 helium throughou heae th t t source temperature range othen I . r wordse th , mercury serves also as a heat transport medium between the helium and water. The efficiency of the mercury topping cycle is calculated to be 8.5%, when 50°C temperature difference between the helium and mercury is assumed. This makes the binary cycle efficiency 47%, as compared with 42% of the conventional Rankine PCS (of Fig. 4a). Notice thaRankine th t e steam cycle assume binare th r ydfo cyclf eo Fig. 4b is of the LMMHD type. [15] This cycle is characterized [15] by an isothermal expansion and by the use of supercritical steam.

5.4.2 LMMHD Ericsson PCS "puree Th " Ericsson cycle describe Secn di .t particularl 3.3.no s i 2 y attractive, as it is, for conventional HTGRs because the cycle efficiency drops significantl temperature th s ya e rang energr efo y delivery increases. There are, however, a number of possibilities for better matching the HTGR energy source characteristics wit LMMHe hth cycls Dga e PCSe On . possibility is to superimpose a number of classical Ericsson cycles, each drive relativela y nb y small helium temperature drop ordee (sayth f f o ro , 100'C). A combinatio Ericsson a f o n n bottoming a wet-vapocycl d an e r homogenous topping cycl makn eca gooea d matc HTGRso ht inlee th , t helium temperature of which is in the vicinity of 550°C. The efficiency of the resulting binary system is estimated at 47%. LMMHe Th D alsEricsson ca o makS nPC verea y attractive matco ht special purpose reactors graphite-moderated,e sucth s ha , heat-pipe cooled reactor propose Refn . di Suc 16 . reactoha r deliver s heasit t almosa t a t constant temperatur provided ean heasa t source perfectly suitabl heatinr efo g the LM of the PCS. With a heat source temperature of 750°C (950°C), an efficiency of 46% (53%) is expected.

363 3.4.3 A dual LMMHD - GTPCS This cycle, shown in Fig. 5, is similar to the dual cycle LMMHD system propose shir dfo p propulsio Refn i , wit exceptione .17 hon carriet :i s out part of the regenerative heat exchange between the gas expansion in the MHD generator and the turbine, whereas in Ref. 17 the gas turbine immediatel generatoryD follow MH arrangemen r e sth Ou . t ease desige sth n of the GT, as it limits the helium admission temperature to ~750°C. The relativel admissios ga w ylo n temperature also reducevapoM L re contensth n ti the helium enterin turbinee g th pressure Th . e drop acros turbine sth s ei adjusteprovido t s a powe e o eds th r requiremen turbocompressore th f to . heliue Th m temperature drop turbin s acrosga e seth makes this duas ga l cycle more compatible, thermodynamically, with HTGRs than the ideal Ericsson cycle neee turbomachinerf Th d.o y can, however disadvantaga e ,b e for certain applications.

~~Liquid M*totHTN R P. MHD G«n. Mi «o rI 2-Pha»«MHO G«n. ^ ^ ~/|S«parator~~

~~Regenerativ« HX A~~

~~N-S1agc Compr«ttorW Intottog«~~

~~N- Stag« Cocnpr»uor~~

~~schematiA Fig5 . c layou duaa f to l LMMHD-GT modified Ericsson cycle and the associated T-S diagram~~

3.4.4 Attainable Efficiencies Table II illustrates the efficiency attainable from selected combinations of HTGR LMMHd san D PCSs. These efficiencies pertai followine th o nt g assumptions generatoD : efficiencMH maximumd e an rth f o y outlet void

~~364 Table IL Efficiencies of Selected LMMHD PCS - HTGR Combinations~~

~~HTGR EfficiencS PC y% Met/Outlet POQ Temp (°C) Absolute % of ideal~~

Reference PCSs 1. 250/750 Conventional Rankine steam cycle 42 72 2. 500/850 turbins Ga e Bra cycln yto e (HTGR-GT) 42* 3. 500/850 Binary cycle: Brayton topping + organic Rankine bottoming 47* LMMHD PCSs 4. 300/750 LMMHD Rankine Superheated steam 45 75 cycle 5. 300/950 Binary: Hg Wet- Vapor + "4" bottoming 47 74 6. 550/950 Binar" y "5 cycle f o s ,a 49 70 7. 550/950 Binary: Hg Wet- Vapor + Ericsson 47 68 8. 550/950 Dual: LMMHD Ericsson cycle wit beloT hG w 750'C 47 68 LMMHD PCS - Special Application 9. 700/750 Heat-pipe reactor with Ericsson 46 67

* This is a net plant efficiency; it accounts for the HTGR coolant circulation power requirement. This power requirement corresponds to efficiency(absoluteS % 2 1PC %o e t th f .)o

fractio e 0.85ar n s turbin;ga d compressoan e r efficiencie e 88%ar s ; effectiveness of regenerative heat exchangers is 97%; friction losses in PCS takee ar n into account auxiliart bu , y power HTGneede th notf e sRo e ar ;th electricity generated by the LMMHD PCS is DC.

5 Discussio3. n All the LMMHD PCSs considered in this work offer, in addition to improved efficiency, a safety attribute - either elimination of the water from the HTGR plant altogether or, at least, separation of the water from the primary helium (and, therefore, graphite) syste liquia my b d metal. Thuse ,th e LMMHth f o De us energy conversion technolog n eliminateca y r o , significantly reduce watee ,th r ingress accident probability.

365 othey LMMHe e th th ma n f rs o handO S DhandlinM e PC L th , e th f go pose maintenance and environmental difficulties. For example, being toxic under certain conditions, mercury mus handlee tb d with special care. Alkaline materials, on the other hand, are destructive to graphite, so that the LMMHD - S shoulPC designee db havo dt possibilito en r liquid-metayfo l ingress accident. otheo Tw r design difficultie associatee ar s d wit LMMHe hth - S DPC compatibility with structural material elevatet sa d temperatures neee th o dt d an , liquie th o dt statM brinL e e beforgth e startin systeme gth .

3.6 Conclusions recentle Th y conceived LMMHD component cycled san s opew nne interesting possibilitie conversioe th r sfo f HTGno R energy into electricity, including: (1) Designing conventional HTGRs (i.e., HTGRs designe o havdt a e helium outlet temperature of 75CTC) to have a PCS efficiency of about 45%. This corresponds to a net HTGR efficiency of ~43% versus <40% with conventional energy conversion technology. (2) Makin hige g th efficienh f o temperatur e us t e operation-abilitf o y HTGR generatinr sfo g electricity might ;i possible tb desigeo t n HTGRs providin efficienciet gne s approaching 47%. ) (3 Attaining efficiencie eved higan s n a s, somewhahas t exceeding% 70 , e ideaoth f l efficiency pertainin e HTGth o t Rg heat source characteristics. This is due to the versatility of the LMMHD PCSs and their flexibility in matching the HTGR heat source characteristic and due to the relatively high efficiency of the LMMHD PCS components. (4) Designing HTGR free f b rotatin o eo st g machinery, promising high reliability and low maintenance systems suitable for unattended applications. ) (5 Reducin ris e water gth kfo r ingress accidents.

366 The incorporation of a liquid metal system in an HTGR power plant introduces, however, a number of safety and maintenance difficulties. Thus, an assessment of the promise of the LMMHD energy conversion technology require comprehensivsa e techno-économie feasibility study e vien th I . f wo potential gai efficiencn ni y offereLMMHe th y b d D energy conversion technology, it is recommended that such a thorough feasibility study be undertaken.

4. REFERENCES [1] GCRA, Evaluation of Small Modular HTGRs Applied to Electricity Generation s CooleGa , d Reactor Associates Report GCRA 84-002 (1984). [2] BREITBACH, G., et al., On the Accident Behaviour of the HTR-Module - A Trend Analysis, Julich Nuclear Research Center Special Report Jul-Spez-260 (1984). [3] SILADY, F.A., McDONALD, C.F., WEICHT, U.H., Trans. Am. Nucl. Soc (19847 4 . ) 290. [4] LOHNERT, G.H., PFLASTERER, G.R., Trans Nucl. Am . . Soc1 4 . (1984) 287. [5] HUBEL, H.J., LOHNERT, G.H., Atomkernenergie - Kerntechnik 47 (1985) 159. [6] GCRA, An Assessment of the Interatom/KWU Modular HTGR Concept, Gas-Cooled Reactor Associates Report (1985). [7] Capital Investment Costs of Nuclear Power Plants (Proc. Techn. Comm. Vienna, 1978 Woite. G ; , Ed.) IAEA Vienna (1975), ] [8 GCRA, Advanced Application Assessmen SUMMARYt- Cooles ,Ga d Reactor Associates, Ref. GCRA 83-010 (1983). [9] QUADE, R.N., RAU, R., HTGR Application for Shale Oil Recovery,

~~GA Technologies Rep. GA-A17081 (1983).~~

367 [10] WADEKAMPER, J.R t al.e . , "Applicatio Hige th hf n o Temperatur e Gas Cooled Reactor to Oil Shale Recovery," (Proc. 16th Oil Shale Symposium), Colorado (1983) 315. [11] BHATTACHARYGA, A.T. e InvestigatioTh , r Attaininfo n e th g Optimal Yield of Oil Shale by Integrating High Temperature Reactors, Kernforschungsanlage Rep. Jul-1905 (1984). [12] ADAR , "CouplinJ. , Standarf go d Condensing Nuclear Power Stations to Horizontal Aluminium Tubes Multieffect Distillation Plants," Advances in Israel Research on Water Desalination (Proc. Int. Desalting and Environmental Assocation, Mexico City, 1976). [13] MELESE, G. and KATZ, R., "Thermal and Flow Design of Helium-Cooled Reactors," American Nuclear Society, 1984. [14] PETRICK , BRANOVERM. , , "LiquiH. , d MetaPoweD MH lr Generation - its Evolution and Status," in Single and Multiphase Flows in an Electromagnetic Field, H. Branover, P.S. Lykoudis and M. Mond, eds. Progres Astronauticn si Aeronauticsd san (19850 10 , ) 371. [15] BLUMENAU t al.e . ,L , "Liquid PoweMetaD MH lr Conversion Systems with Conventional and Nuclear Heat Sources," Proc. 24th Symp. on Engineering Aspects of Magnetohydrodynamics, Butte, Montana, June 24-27 (1986). [16] KIRCHNER, W.L., MEIER, K.L., "Low Power Unattended Defence Reactor, Proc. 19th EECEC, San Francisco, California, Aug. 19-24 (1984). [17] PIERSON, E.S., DAUZVARDIS, P.V., Study of Compact Liquid-Metal MHD Systems Coupled to Helium Gas Cooled Reactors , Argonne National Laboratory Report ANL/ENG-75-01 (1975).

368 GAS-COOLED REACTOR TECHNOLOGY (Sessio) nF OVERVIE MHTGS U F WO R BASE TECHNOLOGY DEVELOPMENT PROGRAM*

~~J.E. JONES, Jr., P.R. KASTEN Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States of America~~

~~Abstract~~

e BasTh e Technology Development Modulae Prograth r rfo m High Tempera- ture Gas Cooled Reactor (MHTGR) provides basic information on fuel performance, fission product retention capabilities, and material properties; the program plan for obtaining such information is summarized, considering the specific needs of the MHTGR concept. Continued international cooperation in HTGR base technology development will contribute to the planned program.

~~Introduction~~

At this time, existing technology is sufficient to show the feasibility of the MHTGR concept and to develop the conceptual design. The base technology development program has been established in direct response to specific design data needs and licensing requirements for the MHTGR concept.

The base technology program is planned to develop the basic correlations concerning the behavior of physical and chemical phenomena; they also provide experimental validation of mathematical models predicting system behavior necessary for licensing (e.g., retention of fission products). The specific development areas discussed here are (1) fuels performance and fission product transport ) graphit(2 , e behavior ) metal(3 d s an ,behavior .

~~Component development requirements have not yet been specifically defined. These development requirements are not included in the base technology program.~~

* This paper was written by a contractor of the US Government under Contract No. DE-AC05-84OR21400.

~~371 Fuel and Fission Products Technology~~

Fission product retention within the fuel coatings is the key ingre- dient to achieving passive safety in the MHTGR, and this area will be emphasized. Figure 1 illustrates a Triso-coated fuel particle used in the MHTGRe fue Th d UC2l. an kernea 2 mixtur, termeUÛ s i lf o ed uranium oxycarbide; it is surrounded by a porous layer (or buffer layer) of pyrolytic carbon which provides expansion volume. Surrounding the buffer volum e successivar e e layer f pyrolytio s c carbon, silicon carbided an , pyrolytic carbon, which together contai e fissioth n n products. These coated fuel particles have very high fission product retention capability, eve t higa n h temperatures w welno ls i establishe t I . d that high-quality coated fuel particles retain essentiall l fissioal y n products under anticipated reactor temperature conditions, as indicated in Figure 2; this strongly support e feasibilitth s MHTGe th Rf o yconcept .

~~OUTER ISOTROPIC PYROUTIC CARBON~~

~~SILICON CARBIDE BARRIER COATING TRISO COATING INNER ISOTROPIC PYROLYTIC CARBON~~

~~POROUS PYROLYTIC CARBON BUFFER~~

~~FUEL URANIUM OXYCARBIDE 'KERNEL~~

~~FIG. 1. TRISO coated fuel particles retain fission products at high temperatures.~~

~~I.U ———————————————————————— 0-^~~

~~0 0.8 0~~

~~o 0.6 o I FRACTIO N es o r> 0.4 1 0~~

0.2 NORMAL o DESIGN LIMIT PEAK FUEL /TEMPERATUR° E TEMPERATURE / oOooo8 0 / i »0000000000000 ' 1000 1200 1400 1600 1800 2000 2200 2400 26 TEMPERATURE (°C) FIG. 2. Coated fuel particles maintain integrity at high temperatures.

372 e samAth te time, durin e coatinth g g process some very small fraction of defective coated fuel particle producede b y sma . Process developmens i t planned to limit defective coatings in fuel to a very low value either by remova f defectivo l e coate y reducinb d r particle o numbee e th gs f r o r pe s defective particles produced in the process itself. Preliminary indica- tion e thaar st process developmen n effectivelca t y reduc e numbeth e f o r defective particles. However, defective particles will not be completely eliminate practicen i de technolog th d an , y development program will address the performance of coated fuel particles having defective coatings to determine fission product release from such particles as well as to guide the requirements associated with process development o obtaiT . n such information, purposely defective coated fuel particle e plannee ar sb o t d irradiate n capsuli d e experiment Hige th h n i Flus x Isotope Reactor (HFIR) k RidgOa e ath tNationa l Laboratory (ORNL), with on-line monitorinf o g fissio s releasga n witd an eh subsequent postirradiation examinationse Th . latter examinations determine the retention of fission products within the coated particles.

Experiments are also ongoing relative to determining the influence of water ingress on hydrolysis of fuel particles having defective coatings. Figur 3 showe n irradiatioa s n capsule use o test d t MHTGR coated particles and fuel rods. These irradiation reflectoe th e carrien i ar s t r ou dregio n of the HFIR at ORNL, as illustrated in Figure 4. As shown, the fuel is located withi graphite nth e cylinder portio capsulee th f no .

Coated fuel particles have to be fabricated into fuel rods, as illustrated in Figure 5. This step involves fabrication processes which n lea ca o somt d e breakag f fueo e l particle coatings e fueTh .l technology development program concentrate ) developin(1 n so g economic processes which maintain the integrity of the fuel particle coatings, and (2) performance testin f fueo g l rods under normal reactor condition s wel a ss unde a l r postulated accident conditions.

Fuel coating process development involves fabricating coated particles which hav higa e h degre f sphericito e producind an y g coating layers having uniform properties. Process development associated with fabrication of fuel rods involves development of processes which minimize the likelihood f breakag o e particlth f o e e d coatingsfabricatioro e Th . n step involves injecting pitc coatef o h d intdbe particlea o s plus matrix graphites i t I .

~~373 Fue3 l Rod r Bodspe y~~

~~H-451~~

~~t t;/22^ s+ o o o / t:' o ° t «°« / o Ä o p/ ô°Qô • 0 O „o c o o o ; - 0 «~~

~~' °o8 '.^J^y ^~~

~~Graphite Fuel Body (8 Bodie r Capsulespe ) (D~~

~~• Capsule~~

~~FIG. 3. Irradiation capsule HRB-21 (USDOE).~~

~~MS-3~~

~~FIG . Irradiatio4 . n wilconductee b l HFIe th Rdn i removable beryllium (RB) facility.~~

~~374 FISSILE (U-235)~~

~~FERTILE (Th-232)~~

~~FUEL PARTICLES FUED LRO FUEL ELEMENT~~

~~FIG . Fue.5 l components.~~

planned that the performance of the fabricated fuel rods will be obtained by irradiation testin capsulen i gHFIe ORNLt th a R n i s; these capsules were illustrated previously in Figure 3. The fuel rods will be irradiated under nominal reactor operating conditions e fueth ;l capsules wil e instrub l - mented with thermocouples, and a sweep gas will be used to determine fissio s releasga n e during irradiation. Following irradiations i t i , planned that the fuel rods undergo postirradiation examination in the High Radiation Level Evaluation Laboratory (HRLEL) (Figure 6) to determine the integrit e coatingth f o y s relativ o retentiot e f fissioo n n productsn I . addition, fuel will be heated to high temperatures simulating postulated accident conditions e performancth d an , f suco e h fuel r retaininfo s g fission products wil e obtaineb l d usin e irradiateth g d microsphere gamma analyzer (IMGA) and postirradiation gas analyzer (PGA) equipment at the Oak Ridge National Laboratory. Such equipment examines coated particled an s determines the fission product inventory still remaining within particles; a stable fission produc a referencs use i ts a d o givt e e quantitative informatio n fissioo n n product retentio f fueo n l coatings. Large numbers of particles are examined to give statistically meaningful information. Figure 7 illustrates the IMGA system. As shown, a singularizer picks up individual particles and a sample changer indexes the particles to a detector wher e gamm th y espectru ra a s measuredi m . Depending upoe th n result f gammo s a analysis e examineth , d particl places i eparticulaa n i d r compartment. Various classification compartment e utilizedb n ca s , with particle separation based on predetermined criteria.

~~375 W -o~~

FIG. 6 . AUTOMATIC PARTICLE HANDLING MECHANISE TH R MFO IRRAOiATED-MiCROSPHERE GAMMA ANALYZER (IMGA) SYSTEM. (AS SINGULARIZER PICK INDIVIDUAP SU L PARTICLE ) SAMPL(8 , E CHANGER INDEXES PARTJCLE TO Ge(Lt) DETECTOR !«JT SHOWN); (C) CLASSIFICATION COMPARTMENTS WHERE PARTICLE IS PLACED IN ONE OF TWENTY COMPARTMENTS DEPENDIN PREPROGRAMMEN GO D CRITERIA,

~~FIG. 7. Irradiated microsphere gamma analyzer for MHTGR.~~

~~Figur A equipmen PG 8 showe e th s t e operateb whic n ca h d eithe root a r m temperature or up to 2000°C. The PGA measures (by a mass spectrometer) the gases released after particle breakage.~~

~~The heavy metal contamination of the fabricated fuel rod is specified % confidenc e~~

A small amount of uranium contamination will be present in the fuel rod matrix material as a result of the coating process. Further, a small fractio f coatino n g defect e expectear s o occut d r durin e coatinth g g process. As a result, fission product behavior within the reactor system mus quantifiee tb orden i d assuro t r e meetin usee th gr requirement f "noo s - evacuation planning." Fission products releasee b n d ca fro e fued th mro l

~~377 TËHWEUâïURt e«St«£SS Aftg US£M »~~

~~CHAMBER CAR »E HEATED T6 20ÖÖ*C «FORE 7HÏ WfcTICU IS 8RÖKEH (PHOTO R-75SÎÏ).~~

~~FIG. 8. Postirradiation gamma analyzer equipment for MHTGR fuel.~~

transported to the heat exchanger system where "plateout" tends to occur on the cooler surfaces. Figur 9 showe e relativth s e locatio e steath mf o n generato e reactoth d ran r core. Ove a perior f timeo d , fission products will "build up" on the steam generator surfaces. During rapid depressurization, some of these fission products could be released to the atmosphere. Present knowledge indicates there is sufficient retention of fission products during rapid depressurization t morbu ,e quantitative knowledgs i e needed A definitiv. e progra bees mha n develope demonstratinr fo d g specific behavior and for validating the mathematical models used to predict fission product transport behavior. More specifically s plannei t i ,o measur t d e the diffusion coefficients of fission products in coatings and in graphite moderator under various reactor conditions. Also, the sorptivity of fission product n materialo s s suc s graphita h e moderato d metallian r c surfaces will be determined for various conditions. In addition, studies of fission-product "plateout" (i.e., without dusd withouan t t water vapor present witd ,an h duswitd tan h water vapor present plannee )ar d along with fission-product "liftoff" experiments under reactor-depressurization

~~378 GRADE LEVEL~~

~~MAII CWCUtATM MOTOI IEACTM CAVITY MAIN HELIUM cooimo CMCUMTOR SYSTEM »CCW~~

~~CROSS OUCI~~

~~SUPERHEATED STEAM~~

~~sHuroowi SHAM GEKERATOR CIICUIATD« VESSEl WQIOI SHAM GENERATOR~~

~~IEACIOI CAVITY - EfOWATER~~

MW(t0 35 FIG. )9 . modula r HTGR elevation. conditions. Both basic datr improvinafo modeline th g f fissiogo n product transpor wels ta s integra la l datvalidatior fo a f systeno m models wile lb obtained. Nuclides to be studied in various tests are iodine, cesium, strontium, and silver.

~~Graphite Technology~~

Graphit n importana s i e t materia e reactoth d n constitutei lan r a s major volume fractioe corth e f o regionn . Figur 9 showe e generath s l locatio e fueleth f do n core, while Figur 0 illustrate1 e s more clearle th y annular characteristic e inne th d oute e an corrd th ran e f reflectoo s r graphite regions. Great Lakes Research Corporation grade H451 graphits eha bee ne fuechoseth l r elemente replaceablfo n th d an s e reflector elements. Stackpole Corporation grade 2020 graphit bees eha npermanen e choseth r nfo t side reflectors and the core support structures. There is much information

~~379 CENTRAL REFLECTOR GIDE REFLECTOR~~

~~REACTOR VESSEL~~

~~ANNULAR ACTIVE CORE CORE BARREL~~

~~CONTROD LRO CHANNELS 16 INNER OUTER4 2 ) SEISMIC KEYS~~

~~COOLANT INLET RSS CHANNELS |12) CHANNELS~~

~~BORONATED PINS~~

~~MW(t0 35 FIG . ) modula10 . r reactor core cross section.~~

on graphite behavior in gas-cooled reactors which generally supports the feasibility of the MHTGR concept. At the same time, additional physical and chemical property data is needed to support the detailed design of graphite component e MHTGth Rr fo undes r normal operation conditiond an s under postulated accidents e graphitTh . e technology development program will obtain the needed data base and involves evaluating variations in graphite strength, determining irradiation creep behavior (an important mechanism for relieving stresses), and reducing oxidation characteristics. e graphitth Mos f o t e technology program effort s relatei - o obtainint d g property data require r reliablfo d e component desig d reducinan n g associated investment risk.

Most of the planned graphite property studies involve irradiation- creep experiments, fracture-mechanics studies, and corrosion/oxidation behavior. Irradiation-induced cree d straian p n date particularlar a y needed since creep strongly influences stresse d alsan so affects graphite properties. Graphite fracture-mechanics behavior tends to deteriorate at high neutron fluences d experimentaan , l creep data will determine graphite reflector lifetimes e fracture-mechaniTh . c studies e initiarelatth o t e- tion, and growth of cracks and provide a basis for understanding the phenomen r f developinfracturo fo a d an é g graphite failure criteria. Relativ o steat e m oxidatio f graphiteo n , both surfac d internaan e l oxidation behavior are planned to be determined.

380 Since they are not pure materials, graphites by their very nature have existing crack flawr so smaterialw whicra inducee e ar hth sy y b db use d dan the manufacturing process. Because scatteth f n measureo i er d graphite property data, and because of insufficient design margin associated with the lower e measurevalueth f o s d properties s plannei t i o ,providt d e graphite property behavio d materiaan r l failure criteri n statisticalli a y meaningful ways. This requires large data bases.

~~Metals Technology~~

Metals technology is sufficiently known to assure the feasibility of metal components in the MHTGR. Additional information needs relate primarily to property data bases required for reliable component design, licensing, and reducing associated investment risk. A metals technology development program has been identified which will provide the required material property correlations e pertinenTh . t component e includear s n i d Figure 11 and consist of the reactor vessel (constructed of SA 533B steel,

~~CONTRO DRIVED LRO / REFUELING PENETRATIONS~~

~~STEEL REACTOR VESSEL MAIN CIRCULATOR ANNULAR REACTOR CORE~~

~~STEAM SHUTDOWN GENERATOR MEAT EXCHANGER VESSEL~~

~~SHAM OUTLET SHUTDOWN CIRCULATOR STEAM GENERATOR~~

~~FEEDWATEfl INLET~~

~~FIG. 11. Modular HTGR. Side-by-side prismatic fuel.~~

381 which is a Light-Water Reactor pressure-vessel steel); the duct which connects the two vessels (made of Alloy 800H); the core lateral restraint system (includes the core barrel) surrounding the graphite core (made of 2.25Cr-lMo and Alloy 800H); and the steam generator (made of Alloy 800H and 2.25Cr-lMo steel). Long-term testing is needed to complete the materials property data base ; key measurements involve fracture mechanics, mechanical properties, radiation effects, environmental and corrosion effects, and weld properties. These measurements on material properties are planned in orde provido t r e assurance that components will perform satisfactorily over their design lifetime. Further, materials are to be qualified under an ASME Code Case specifying data base requirements. Sufficient data wile lb obtained so that rational extrapolations of materials performance can b!e made into areas where dat s limitedi a ; tha, datis ta trends wile b l established.

Overall, information will also be obtained which provides material property data importan e desigth componentsf o no t r exampleFo . , thermal agin n lea changeo gca t d materian si l ductilit d creeyan p strengt d thihan s ha a direcs t influenc n desigo e n requirements. Additional information o n mechanical properties as a function of environment and aging are to be obtained for both 2.25Cr-lMo and for Alloy 800H. In particular, emphasis will be placed on understanding the influence of aging effects on material properties since such information reduces investment risk associated with component performance.

A significant portion of the Metals Technology Development Plan is associated with high temperature fracture mechanics studies. Fracture mechanics involves basic measurements associated with the initiation, propagation d arresan , n particularf cracki o t d an s , investigates conditions associated with crack growth and crack arrest. Such studies are believed to be a key to understanding why materials fail; this understanding is particularly important in high temperature applications of materials. Such studies are complicated by the influence of factors such as agin d environmenan g fractur, on t e mechanics behavior s plannei t I d. that significantly more work in the above areas be carried out on 2.25Cr-lMo and Alloy 800H, including decarburization studies. Irradiation testin f Alloo g A y533S 800Bd an steelH s includei e splanne th n i dd program. Also e cree th d ,ruptur an p e behavio A 533S Bf e o rstee th t a l temperatures associated with postulated accident conditions wile b l studied. Othe rD beinR& area gf o se planneMetalth n si d Technology

382 CREEP TEST SPECIMENS CREEP TEST ASSEMBLE PROGRESN I D DAN S oOoJ . FIGCree12 . p testing equipmen specimend an t r MHTGsfo R metals. Development Program are irradiation behavior of materials, and the effects of HTGR-He environmen d thermatan l agin materialn go s properties.

e equipmenth Som f o e t being use Metaln i d s Technology Developmene tar illustrate . Figur13 n Figured i 2 showd1 an e 2 a crees1 s p test assembly being use o test d t 2.25Cr-lM d Alloan o y 800H, along with creep test specimens. Figure 13 shows some lever-arm test units for material creep tests at high temperatures in an HTGR-He environment.

~~FIG. 13. Lever-arm creep test equipment for MHTGR metals at high temperature.~~

~~International Cooperation and Base Technology~~

e UniteTh dFederae Stateth d lan s Republi Germanf o c y presently carry out a cooperation under an Umbrella Agreement covering MHTGR technology development. Also, the United States and Japan are cooperating in MHTGR researc d developmentan h e U.S./FRTh . G cooperation involve a sub ) - (1 s progra n metali m s development which addresses eight specific studied an s (2) a subprogram in fuels, fission products, and graphite which covers

384 sixteen specific investigations. The U.S./Japan cooperation involves testin f MHTGo g R fuels including their performanc t higa e h temperatures; further cooperatio s plannee graphiti n th n i d beins ei ared g an apropose d in the metals area. These cooperative agreements contribute to expand base technology information pertinent to the MHTGR. Continuing and expanding such cooperatio n hav ca nbeneficiaa e l effec n increasino t amoune th g f o t property-data information and limiting development costs.

~~Conclusions~~

Base Technology Development for the Modular High-Temperature Gas- Cooled Reactor (MHTGR) provides basic informatio n fueo n l performance, fission product retention capabilities d materiaan , l properties. Present information is sufficient to show the feasibility of the MHTGR concept, and to develo e conceptuath p l design e basTh .e technology development program is based on the specific design data needs of the MHTGR concept; these programs will develo e datth pa base d validatan s e systeth e m response models require r detailefo d dr licensingdesigfo d e plannean nTh . d base technology program is relatively modest and is well focused on the needs of the MHTGR concept. Continued emphasi n internationaso l cooperatio basn i n e technology development will increas e amounth ef properto t y behavior informatio wels n a s limila t development costs.

~~^385 STATU DEVELOPMEND SAN GERMAE TH F TO N MATERIALS PROGRAMM HTGE TH RR EFO~~

~~H. NICKEL, F. SCHUBERT, H. SCHUSTER Institute for Reactor Materials, Kernforschungsanlage Jülich GmbH, Jülich, Federal Republic of Germany~~

Abstract During the last decade the research and development of material e HTG n Germanth i R r fo sy have mainly been directed towards the qualification of high temperature structural alloys for heat transferring components of advanced nuclear process heat plants. The availability of Fe/Ni-base and Ni-base alloys for long term service loading has in principle been established. Furthermore, materials and optimization have indicated that improvements in the corrosion properties are also achievable. The materiastatue th f so l evaluatio optimizatiod an n e nar demonstrate do (AlloM wit 2 1 h y o C NiC 6172 2 rd Thermo )an n 497hear 2fo t exchanging components, witsteee WerkN th h lDI - stoff -Nr. r absorbefo 1.498CrNiMoN8 d ) X an ( 16 r1 6 1 b control rod carbod san n fibre reinforced carbon t (CFCho r )fo ducting. The continuing HTR structural materials programme is aimed at improvin e qualitth g d datan ye materiala th bas f o er fo s components of advanced HTGR's, both for nucler heat application and for the steam cycle. Emphasis is given e determinatioth o t - f lono n g term design valuer fo s different semi-finished products including weldments; - to improve service life estimation rules; - to improve constitutive equations for inelastic analysis and to develop design rules; o develot - p fracture mechanics criteri lear fo abefor- k e- break argumentations. e generaf thio Th sm o assur t ai lwor e lons i kth e g term integrity of class I components. Using typical examples, this paper show e e theoreticastatuth th s f o s d experimentaan l l work related to these objectives.

1. Introduction It is intended at this meeting to provide a review of the current recene statuth d ts an progres stechnologe madth n ei y of gas-cooled nuclear reactors. This report shows the status achieve solutioe materiale th th n i df no s problems. Extensive effor e s qualificatioi devoteth t o t d f o n materials, o Germaconcentratetw e n th advance n o d d HTGR concepts, the HTGR steam cycle and HTGR for nuclear process heat.

387 The materials for the helium-cooled HTGR of the pebble bed type have been defined and they are qualified, the manufacturing procedure e requireth r fo sd product forms have been proved /I, 2, 3, 4/ and the semi-finished products and the fabrication method e availablear s . This statement stands withou y reservatioan t e componentth r steaa fo nr mfo s cycle HTGR with working temperatures up to 750 °C as is demonstrate e commissioninth y b d e THTR th e spherica .f Th o g l fuel elementstructurae th d san l graphite have been developed and qualified, although long-time irradiation experiments and test f simulationo s f emergenco s y condition stile ar s l going on /I, 5, 6, 7/. Some restrictions, however, must be made for component a nuclea n i s r process heat plant wite highesth h t working temperature (up to 1000 °C) concerning the permitted service life. The ongoing work for structural materials is concentrate e improvementh o e lont d th g f tero t m reliability of metallic components for nuclear process heat application . 8/ /3, ,7 In this pape e wilw r l concentrat n metallio e c materialr fo s structural components of the steam cycle and process heat HTGR's. 2. Comments on selection of candidate materials For the selection and qualification of metallic materials the following properties have been take accountn i n : - strengt d lonan hg term ductilit r componentfo y s whicn hca be designed with time independent materials data; - creep resistance and microstructural stability for those components which must be designed using time-dependent design values; - corrosion resistance in the working atmospheres; - hot cold-formabilitd -an weldabilityd an y ; - possibility of non-destructive inspection during manufacture and in the plant; - sensitivity to neutron irradiation induced embrittlement (this r onlthosfo y e componente th n si whice ar h core-region). desige th constructiod r nan Fo THTf no R components steele sar usedr whicfo , h experience from fossil-fired plants i s available. Those steels are: e ferritith - c hea te fueresistanth l r fo 3 t o steeM 5 1 l loading tube system; e externath r fo l e stee 0 CrMo1 component2 th 2 X l 1 V - f o s the steam circuit; e creeth p- resistant high temperature alloNiCrAIT0 1 X y i 0 (Allo2 2 3 ysteae th 800 mr )fo generato d livan re steam heade d tubean r s enclosee prestraineth n i d d concrete pressure vessel. - the high temperature stainless steel X 8 CrNiMoNb 16 16 for contro absorbed an l r rods r thi.Fo s steel there exists already experience concernin e effecth gf o neutrot n irradiation from experiment e fassth t don r breadefo e r reactor projects.

388 e cas Ith f componentno e r nucleafo s r process heat plants, creep resistant alloys are needed. From the possible alloys used in conventional petrochemical plants or in gas turbines, the alloy (FIG.) 1 - NiCr22 Col2 Mo (Alloy 617) s e highesselectewa th r tfo d working temperaturesr Fo . component parts with lower temperatures of about 850 °C the alloys - NiCr22 Fel (HASTELLOo 8M ) YX and NiCrAIT0 1 ) 0 (AlloX H 2 0 2 - i3 80 y have been qualified.

~~G Z JC t 500 - «0 s 200- \ \ 1jj > \ \ 100- X ^ > % v \ {\ >• \ a \ v \ \ 50X : \ 1 \ \ \ \ \ ^ o N 1 \\ \ \\ v^v \ IN 738 LC T" 15M03 \ \ S\ \~~

~~. X20CrMoV12A X \ 1 V „ AISI318 / \ 5- X 10NIC TI3I rA 2 20(434)\ (allo\ y 800)~~

~~NICr22 Co12Mo 2- (alloy 617)~~

~~1- 0 75 'lOOO 5(0 ) temperature/°C Fig : Mea1 . n 100,00 h 0cree p rupture strent f somo h e typical high temperature steel alloyd san s~~

As an alternative for Alloy 617 a new Fe-Ni-Cr alloy with about 12 % W has been developed by Thyssen-Edelstahlwerke AG, Krefeld, FRG. This alloy, Thermon 4972, is now also under qualificatio nuclear fo n r components /9/. e welth l f o know The enon stee, material0 CrMo1 2 2 X l1 V s for modern German coal-fired plants, has been improved concernin levee th gtracf o l e element applicatioe th r sfo n ni the THTR steam circuit 710, ll/.

389 To minimize the sensitivity of Alloy 800 for loss of ductility due to ageing at temperatures between 550 and Al+Tcontente d th an limiteds i C C ° f so 0 .65 Meanwhilt i e s beeha n proved that long-term propertie f o this s THTR-version reach the same level of creep rupture strength solutioe th s a n heat treated witH Allo 0 h 80 yhighe r leven i l C and Al+Ti. creeAh compariso0 p1 rupture th f no e strengthse /12th f /o candidate materials shows that Alloy 617 has the potential for application at the highest service temperatures. In the design of heat exchanging components, creep rupture strength was given priority for the selection of the materials, howeveris t I ., importan o t takt e into consideratio e influencth n f corrosioo e e operationath y b n l environment creee th pn s o behaviour . Protection against corrosion in the circuits of an HTGR is achieve e followinth n i d g way: Steam corrosion in the secondary circuit is controlled by conventional means, i. e. appropriate conditioning of the feed water. e heaTh t transferrin e primarth n i y s circuitga g , heliums ,i slightl ubae y th rcontamine n i 2 N y HJDd b d, CH. an CO , 2 H , range. Chromium is the mdtal element governing The corrosion reactions with the Fe- and Ni-base alloys. No deleterious hot s corrosioga n effects have been o temperaturefount p u d f o s 750 °C. Compositions of the impurities with excessively high or verw carbolo y n activitie y causma s e carburizatior o n decarburization, respectively y O keepinB (FIGC e . th 2) .g partial pressure of the cooling gas helium in a well defined range, unacceptable carburization/decarburization reactions

~~Fig. 2; Corrosion areas in the stability diagram for chromium~~

390 can be avoided. This range may already be obtained by the impurity composition which usually is formed without conditionin /13/s .ga e th f go 3. Components for the steam cycle of power generating HTGR For the structures of a steam cycle HTGR the required steels are well specified Germae eitheth n ni r standards e /14ar r /o give ASMn i n E Boiler Code, Code Cas /15/7 .eN4 Nevertheless, in order to take advantage of the materials capacity, refined methods of life time prediction and insurance of long-term integrit componente th f o y requirede sar followine .Th g tasks are handle onginn di g research work: - determin e influencth e f preageino e d colan gd deformation on long-term properties; - establishing life-time fraction rules; - establishing fracture mechanical criteria in the creep-fatigue region. For the THTR components these criteria have been fulfilled in a conservative manner. The ongoing work for the above mentioned areas is not only being carried out in Germany, but it is part of the international cooperation with the United States, Japan and Switzerland /2/. Durin e THTth gR materials e externaprograth r fo ml steam cycle tubes, the postulation of "leak before break" has been proved for the steel X 20 CrMoV 12 1 /II/. Calculations using the fracture mechanics approach predicted satisfactorily that for this stee "leae lth k before break" criterio fulfilles i n d in the component. . 4 Materia contror lfo l rods The control and absorber rods are essential parts for the n HTGa safet Rf o yplant . These metallic componente ar s exposed to irradiation by thermal and fast neutrons. The effec f faso t t neutron n materiao s l propertie limites si o t d the temperature regime up to about 350 °C. At higher irradiation temperatures, however, the helium atoms, produced by an n,ot-reaction with boron and nickel, decrease the ductility of steels and Ni-base alloys. This high temperature embrittlement increases with increasing neutron fluences and test temperature s wela s , wita e duratiole th hth f o n irradiation at high temperature. Emergency conditions in advanced HTGR' leay short-tero sma t d m temperature excursions up to about 850 °C in the control rods. For the advanced HTGR, therefore, the selection of material for the control rod tubes is of crucial importance. In a screening program a number of high temperature steels and Ni-base alloys have been irradiated an t 400a d 0 ,50 werd an e C post-examine° 0 60 tensiln i d short-terd ean m creep tests e comparisoTh . f post-irradiatioo n n tensile rupture elongation (FIG. 3) lead to an optimized version (KA) of

~~391 50-, •\ 873 K 1 = 973 K u l*fl SO rr.. * 38 1 3O 33 25 o 24 19 s 15 f-10- 13 14 o. a 6.8 ö 5- 4.6 3.3~~

~~1.5 1-~~

0.6 0.5- 1 0.4 m- 1 1.4981 1.4981 1.4876 24603 1.4981 1.4981 1.4876 2.4603 KA2 KA2 Fig. 3: Creep rupture strain of various metallic high temperature alloys befor afted ean r neutron os o irradiation ($,. - 3 x 10 m~ ) and short fracture life-time (< 100 hrs) S unirradiated E3 irradiated

1.4981 (X 8 CrNiMoNb 16 16) in comparison with Alloy 800 and HASTELLOY X. By the alloy optimization work /16/ a satisfactory level of post-irradiation rupture strain both in tensile and creep testing could be preserved (FIG. 4). The irradiation tests are continuing in order to accumulate a fluence of 10 cm" , the expected maximum fluence of control rods in HTGR plants. The results confirmed the use of this optimized thermal mechanically treated 1.4981 for future control rods.

5. Materials for intermediate heat exchanger and methane reforming tubes in a nuclear process heat plant The qualificatio s reache ha f Allo o n 7 a statud61 y s which allow o realizt s a desige n timr He/He-IHfo e f 100,00o X 0 a hourmaterial t a d s140,00 an sC ° 0 temperatur0 95 o t p u e hours for steam reformer tubes at working temperatures of up . °C 0 to90 All the basic work concerning - creep rupture testin 30,00o t p gu 0 hours; (FIG5) . - high cycle fatigue testing; cyclw lo e - fatigue (withou witd tan h hold times,; FIG6) . and - influenc ageinf eo short-tern go m properties ) (FIG7 . has been done by the joint effort of the German partners in the HTGR materials working group. So we obtained design curves both for creep and fatigue (FIG. 8), and confirmed the

~~392 Irradiation experlmen! HTR-K1 HTR-K2 HTR-1 HTR-2c HTR-2AB 200~~

~~100 80 . Material «0 1 .4981 -K A optimized u ç M)~~

~~20 a 1.4981~~

~~a creep rupture elongation O hot tensile rupture elongation~~

~~10' 1010 10 ii Thermal neutron (luence Fig : Ruptur4 . e strai stainlesf no s steel 1.4981 (CrNiMoNX8 ) afte16 6 1 br neutron irradiation (Irradiation temperature: 400 °C; Test temperature: 850 °C)~~

~~200- 200- e 1 iooH 100-~~

~~50-~~

~~20-~~

~~10-~~

~~800 «SO 900 «0 1000 TI'CI he air ••••• ...... , • ,,.„•...... • . . 1^. | 10 10 10 10 10 if I h cree | % p ptimt 1 straio et n tim rupturo i t e (h * t e~~

~~Fig: 5 Cree. p propertie f INCONEso 7 61 L~~

~~393 rangr fo e design curves~~

~~10' cycle failuro st e resultF LC r INCONEFig fo s: 6 7 . 61 L~~

aged 1000h a aged 300h 0 aged 100h l treateso o d —i——i——i——i——i— 0 80 100 0 60 0 temperature °C n o fracturFig : C 7 .Influenc° e0 f 90 o ageine t a g elongatio o (AlloC 0 1 yo M NiCf 617 no 2 2 r )

application of the linear life time fraction rule for creep and fatigue vtZ D= C R. where exhaustio= creeo D t e pdu n = actua 9 t ld temperatur an duratio . J < t a nT e" to- = allowable time at d. and temperature T. Kl l l and Df= where exhaustio= f fatiguo D t e du ne N. = actual number of cycles at AÉ. . and T. N^. = allowable number of cycles at /»>£. ana T.

~~394 -7MPa~~

~~Fig. 8: Scheme of creep-fatigue analysis (value INCONEr sfo L 617)~~

The analysis of component life for the high temperature PNP-components cannot be treated by elastic analysis; inelastic analysis is necessary to confirm a realistic service life. The strain range for fatigue is calculated to be less than 0.03 % which can be neglected for life time calculations /17/. r steaFo m reforme re operatinth tubes n i , g timf o e 140,000 hours 160 start-up and shut-down loading cycles are assumed, which give runninsa g5 hour tim87 f seo between shut downs. Taking into consideratio e relaxatioth n f stresseo n s caused by temperature gradients through the tube wall, the life-time calculation gives the following result (FIG. 9) /18/: Assuming that each cycle produces the same amount of consumptio f life-timo n s a doee efirs th s t cyclee th , life-time exhaustion 1 allowfacto = 5 start-u D 45 rs d an p shut-down procedures, more than twice the number postulated operatioe foth r n time. The life of heat exchanger tubes is mainly determined by the primary stress level; thermally induced secondary stress may be treated as unimportant due to the small wall thickness. The estimated primary stress in the tube wall is less than N/m3 o thas msafa t, e operation tim f 100,00o e 0 hourn ca s be achieved extension .A thif no s operation time need svera y carefully e analysitubesth f ,o s which doet seeno sm feasible. The questio importance th f no creef eo p ratcheting effectn so the He/He-heat exchanger components canno e handleb t y b d usin y calculatioan g nASME-Ce giveth n i nC N47 inelastin .A c

~~395 OU J —— — —— ———~~

50 ~~~^*^ S,^^ Relaxation equationluraaxial) ^x>^ W modified retain LOAD SITUATION Sx X &Sf!wi= 0 r i xs^ 30 r. =SSmm' "^\ %î^x^ i ^^ 20 vO \ A T70*1 C ^*^^ ûp-Sbor T =900«( 1 ^"-5^5 ^^ 10 ^ ^ -^ -—•7 - 1 \>idditionot oxic t stress 'of 3 N/mm1 n

~~U.UO 875 0.0rt5 rt|~ T~~

~~^0.04 0^ /" 1 0.03 /- ~t/> /' 0.02 / / 0.01 ^ - " ___.— - 0 0.01 0.1 10 K)0 1000 10(300 t ime h~~

~~Fig. 9: Creep relaxation of thermal stress and resulting strain behaviour at the inner surface of the tube NiCr 22 Co 12 Mo (Alloy 617)~~

analysis of finite element programs leads to very expensive and time consuming computer times. Therefore e degreth , f o e ratcheting due to hot streakes in the helium has been experimentally investigated. It has been found that /19/: - creep ratcheting may occur if the primary stress is low (FIG. 10); - but the accumulated ratcheting strain remains less than the permitted strain limits. t headea heaho f to e r th exchanger r Fo e postulateth , d emergency condition determines the dimensioning. The header must be safe against creep buckling. In the case of a helium/helium heat exchangee postulateth r fo r d emergency condition, loss of pressure in the secondary circuit exposes th0 degreese95 componentt a .r a pressurba Thi o 0 t ss4 f o e exposure may cause a loss of geometrical stability. The judgement of the component behaviour must be done by inelastic analysis, in which the deformation with increasing time mus evaluatede b t .

~~396 20- 2 Cp0/Nmn X10NiCrAlTi3220H~~

~~35 constV * . = 950°C~~

~~sX> • ratcheting ff-~~

~~10- 25~~

~~5- LL ^25 100 200 300 400 time/h FigCree; 10 . p curve under axial primard y an stres) A < s superimposed alternating secondary stress (i.e. ratcheting^»).~~

A parameter stud f creeo y t headepa ho bucklinf e o r th r fo g He/He-heat exchanger has been performed /20V: dimensions were 437.5 mm/12 externa; 5mm l FIGURpressur. °C 7 E bar2 95 e4 = ,T 11 show e increasth s f ovalit o ea functio s a y f timeo n . Accordin o thit g s calculation a ,tota l0 2 exposur= t f o e hours for the starting ovality of ^= 1 % may be tolerated. In reality e temperaturth , e will decrease immediately after the emergency event. If the cooling rate is 5 K/h, failure ducreeo t e p bucklin expectet no s i g d (FIG. 12).

~~4 -i~~

~~INCONEL 617 1HX tube T =950°C 3- R =41 bar~~

~~a> o~~

~~Fig. 11: Time dependence of tube ovality~~

~~397 OK(h -IK/h~~

~~20 30 time [h)~~

~~: Influenc12 Fig. coolinf eo g rat creen eo p buckling~~

. Material7 ductt ho sr sfo r steaFo m cycl e madeb s plantsducen ga ca tfroe mth , sheet of those materials which have already been mentioned, sucs a h a nuclea r fo rHASTELLO t procesBu . X Ys heat plant, with higher temperatures t ducf ho o higt a h, temperature capabilit necessarys i y Germae th n n.I materials programme, CFC-component e undear s r development (FIG. 13) e figurTh . e shows a test of a ring-section. The high temperature evaluation of this material is now in progress. . Futur8 e necessary effort For steam cycle plants, design code concretr sfo e vesseld san graphite components are to be formulated. In respect to codes for metallic components, some additional effor necessarys ti . For nuclear process heat components, especially for the heat exchanging components, there a demanwil e r inelastib l fo d c analysis or simplified inelastic analysis for the structural design. The problems with inelastic analysis which remain to be solved are: 8.1 Constitutive equations Constitutive equations should describ e three-dimensionath e l stress/strai R desigHT n e nbehaviourth e worr th fo k n I . rules, besid e analyticath e l approach give /12n i n22/, ,21 , Norton' e three-dimensionath sn i cree w la p l formulatios i n used. Furthe e applicabilitth r e ORNL-modeth f o y wels la s la

~~398 the Interatom-model for HTGR components is under consideration.~~

~~The main problem with all constitutive equations remains the scattemateriale th f o r s data which usede havb o .et~~

~~Fig. 13: CFC-ring test from a tube with 900 mm 0.~~

399 8.2 Influence of anisotropy of mechanical properties in the semi-finished products A comparison of the creep behaviour of bar material and axially loaded tube specimens (FIG ) show. 14 importanc e sth e of the manufacturing process. In our case the tubes were subjected to a final cold deformation in order to fulfil the tolerances of straightness. The prior cold work lead to a significant change in the stress-time relationship 7237.

~~X: rodspecime n • IHX tube calculation ° : reformer tube~~

~~X 10 NiCrAm 32 20~~

~~0 25 se 180 125 15Q 175 206 225 time / h Fig. 14; Creep curves for axial loaded specimens at 950 °C (é = 30 Nmnf ^)~~

3 Weldment8. s The strain in the areas of a weld seam should be restricted to 50 % of that allowed for the parent metal according ASM N-47C EC wele .Th d itsel vers fi y difficul treao t t with inelastic analysi o different e du s t stress/strain behaviour of its sections (FIG. 15). The figure demonstrates the remaining creep deformation acros a weldmens t after creep exposure at 850 °C of 500 and 1610 hours 7247. The above mentioned limitatioo weldmento e straie th e th b n f i ny no tma restrictive. 4 componente Flaw8. th n si s At ambient temperatur e fracturth e e mechanics calculatios i n used if detectable flaws are present or a postulated flaw in the component is expected. The analysis should result in the proof that within the next inspection period or the total service life cracth e k growth wilt lea no o spontaneoult d s catastrophic fracture. In high temperature fracture mechanics both fatigu d creean e p crack growth mus e examinedb t , these are jusw undeno t r intensive examination n steai r mfa o S . reformer tubes of Alloy 800 H, the fatigue crack growth may be describe y thresholb d Paria d w s(FIGan d la typ ) f .o e16 7257. Creep crack growth evaluatio underways i n .

~~400 20 H NiCr22Co12Mo =85T C ° 0 test duration: 1610h~~

~~15- = 61,a cr 5MP ^-o-test duration 500h~~

~~o. O) eu 10-j~~

~~0 15 30 45 60 75 90 z - position on specimen /mm~~

~~Fig. 15: Axial strain distribution along a circumferential welded tube after axial creep exposure~~

~~IS)~~

~~. _ 25 mm CT-Probe. •CO~~

~~Is RSO-RohRSO-I r • -12, m m 5 CT-Probe ^~~

~~KjACMPa tola' 8 9io'~~

~~: Compariso16 g Fi f o fatigun e crack growth r datfo a T'CT-specimens, l/2"CT-specimens and steam reformer tubes with circumferential defects ) °C 0 NiCrAIT0 85 (X1 , 20 2 3 i~~

401 8.5 Uncertainties in the knowledge of expected loading history In e additiouncertaintieth o t n d scattean s r concerning materials and product properties, the uncertainties in the loading history hinder the exact calculation and prediction of the real life-time of a component.

. Fina9 l comments e problemTh s mentione e dmai th abovn e areaar e f futuro s e efforts. e materiaTh l programmes executed have demonstrated thae th t main materials problems are solved. Continuing efforts are necessar o ensurt y e lonth eg term reliabilit d safetan y f o y the components.

ACKNOWLEDGEMENT The authors thank to all the members of the working group "Arbeitskreis 8 of the EG-HTR 'Materials n i and the "Fachkreis Auslegungskriterien" for the fruitful cooperation. e wor s sponsoreTh i k y b d Bundesministe Forschunr fü r Technologid un g e (BMFT), Bundesminister des Innern (BMI), and Ministerium für Wirtschaft, Mittelstand und Technologie (MWMT) des Landes Nordrhein-Westfalen.

10. References /!/ NICKEL, H.: "Long-Term Testing of HTR Fuel Elements in the Federal Republic of Germany" KFA-JÜ1.-Spez. 376, 1986 /2/ "Special Issues of Nuclear Technology on "High Temper- atur Cooles eGa d Reactor Materials" Post. G ed . Wirtz. .R ,K Nickel. ,H Rittenhouse. L . ,P , T. Kondo, 2 (19843 70 - - l 1 ) 1 . Nr , 66 / "Metallisch/3 e Werkstoff HTRr efü " Editor: Der Minister für Wirtschaft, Mittelstand und Verkeh s Landede r s Nordrhein-Westfalens, , Ban14 d Schriftenreihe "Energiepolitik in Nordrhein-Westfalen", 1982

402 A/ "HTR-Komponenten" Editor: Der Minister für Wirtschaft, Mittelstand und Verkeh s Landede r s Nordrhein-Westfalens, Band 16-1, 16-2, Schriftenreihe "Energiepoliti Nordrhein-Westn i k - falen", 1984 /5/ Special Issue of Nuclear Technology on "Coated Particles", 35, No. 2, (1977) 205-573, ed. R. G. Post, K. Wirtz, T. D. Gulden, H. Nickel /6/ SCHULZE, R.-E.; SCHULZE, H. A.; RIND, W.: "Graphitische Matrixwerkstoff r fü kugelförmige e HTR-Brennelemente, Ergebnisse der Materialentwicklung Bestrahlungserprobung"r unde d , KFA-Bericht JÜ1.-1752, 1981 /?/ NICKEL, H.; SCHUBERT, F.; SCHUSTER, H.: "Evaluatio f Alloyo n r Advancefo s d High Temperature Reactor Systems", Nucl .Desigd Engan .8 (1984) n7 1 ,25 - 265 /8/ NICKEL, H.; SCHUBERT, F.; SCHUSTER, H.: "Very High Temperature Design Criteri Nuclear fo a r Heat Exchanger in Advanced High Temperature Reactors" d Japanese/Germao3r n n Joint Semina n Researco r d an h Structural Strengt d NDE-Probleman h n Nucleai s r Eng., Stuttgart, 1985, Nucl. Eng. and Design, to be published (1986) /9/ HUCHTEMANN, B.: "Beitrag über den Einfluß der Elemente Mangan, Sili- cium, Chrom, Niob und Wolfram in FeNiCrW-Legierungen f reaktionskinetischau e Vorgängr Phasengrenzde n a e e Gas/Metal n i dosiertel m Heliu i Temperaturebe m n " °C 0 95 zwisched un 0 n75 Diss., RWTH-Aachen, 1984 /10/ TOLKSDORF, E.; SCHNEIDER, K.; GRÜNLING, H. W.: "Toughness Behaviou e , SteeCrMo0 1 2 th 2 X 1 lVf o r Result Optimizinf so Basee Welth gd -an d Metal" FRACTURE PREVENTION AND AVAILABILITY, 9. MPA-Seminar, Stuttgart, 1983, Staatliche Materialprüfungsanstalt Universität Stuttgart

/11/ BERNECKER GNIRSS; ,G. SCHULZE; ,G. , H.-D.; JULISCH: ,P. e Fractur"Prooth f o f e Resistance e Pipeth th n i sf o e THTR-300 Plant Using High Temperature Burst Tests with CrMo0 Components1 2 X 2 1 V " FRACTURE PREVENTION AND AVAILABILITY, 9. MPA-Seminar, Stuttgart, 1983, Staatliche Materialprüfungsanstalt Universität Stuttgart /12/ NICKEL, H.; SCHUBERT, F.; PENKALLA, H.-J.; OVER, H. H.: "Material-Mechanics View on the State of Research Concernin Desige th Basie High-Temperaturf r th gno s fo e Components in HTR-Plants"

403 FRACTURE BEHAVIOU D NONDESTRUCTIVAN R E TESTING. 8 , MPA-Seminar, Stuttgart, 1982, Staatliche Material- prüfungsanstalt Universität Stuttgart /13/ QUADAKKERS, W. J.; SCHUSTER, H.: "Corrosion of High Temperature Alloys in the Primary Circuit of High Temperature Gas Cooled Reactors" Part 1: Theoretical Background, Werkstoff Korrosiod eun 0 15 6 (1985) n3 - 1 ,14 QUADAKKERS, W. J.: Par : Experimentat2 l Results, Werkstoff Korrosiod eun 7 6 (1985)34 n3 - 5 ,33 /14/ VDTÜV-Blätter (VDTÜV-Essen), TRD-Vorschriften und AD-Merkblätter (Carl Heymanns Verlag, Köl Berlin)/ n , DIN-Vorschriften (Beuth Verlag Berlin) /15/ ASME-Code, Case N 47-17, "Class I Component in Elevated Temperature Service, Divisio, l" n ASME (1979) /16/ THIELE SCHUBERT; A. . ,B NICKEL; ,F. : ,H. "Beitrag zum Hochtemperaturverformungsverhalten des austenitischen Stahls 1.4981 nach Neutronenbestrahlung" KFA-JÜ1.-1904, März 1984 /17/ SCHUBERT, F.; BODMANN, E.: "Versagenssichere Bemessung von Bauteilen im Hochtem- peratur-Reaktorbau" Beitra r DGM-Diskussionstagunzu g s Fachausschussede g s "Werkstoffverhalten unter mechanischer Beanspruchung" am 18.719. Sept. 1986 in Karlsruhe

/18/ BIENIUSSA ; BREITBACHK. , . PENKALLA; H OVER G. ,. H , , H.-J.; SEEHAFER, H.-J.: "Life-Time and Creep Ratcheting Calculation of Two Typical HTR-Components" 2nd International Semina n "Standardo r Structurad an s l Analysi n i Elevates d Temperature Applicationr fo s Reactor Technology", Venice, 14-17 Oct. 1986 /19/ ZOTTMAIER, R.; OVER, H. H.; SCHUBERT, F.; NICKEL, H.: "Untersuchungen zum Kriechratchetingverhalten von Rohr- proben" KFA-JÜ1.-2019, Sept. 1985 /20/ "Auslegungskriterien für hochtemperaturbelastete metal- lische und keramische Komponenten sowie des Spannbeton- reaktordruckbehälters zukünftiger HTR-Anlagen" KFA-Jul.-Spez.-347, Febr. 1986 /2l/ PENKALLA, H.-J.; ROEDIG, M.; HOFFMANN, M.: "Transferability of Design Values and Constitutive Equations on Multiaxial Loading Conditions" SAFET D RELIABILITAN Y F PRESSURO Y E COMPONENTS with special emphasis on FRACTURE EXLUSIONS, 10. MPA-Semi- nar, Stuttgart, 1984, Staatliche Materialprüfungsan- stalt Universität Stuttgart

404 7227 SCHUBERT, F.; SCHUSTER, H.; DIEHL, H.; HONEFF, H.: "Zeitstandverhalten und Gefüge des Werkstoffs Nr. 2.4663" Presentatio MWMV-VDM-Seminart na , Werdohl, 23.01.1985 7237 SCHUBERT ROEDIG; ,F. NICKEL; ,M. : ,H. "Multiaxial Loading Tests of High Temperature Reactor Components" SMIRT 8-Post Conf., Post-Semina , Paris6 . ,No r August 1985 7247 KAFFKA, M.; ENNIS, P. J.; SCHUSTER, H.; NICKEL, H.: "Beitra Zeitstandverhaltem zu g n artgleicher Schweißver- bindungen einer Nickelbasislegierung unter besonderer Berücksichtigung des lokalen Verformungsverhaltens" KFA-JÜ1.-2092, Oct. 1986 7257 R. J. Kwasny: "Bruchmechanische Untersuchunge n hochwarmfestea n - Le n gierunge m Temperaturbereici n n Raumtemperatuvo h s bi r " °C 0 90 Dissert. RWTH Aachen, 1986

~~405 KVK AND STATUS OF THE HIGH TEMPERATURE COMPONENT DEVELOPMENT~~

~~W. JANSING, H. BREITLING, R. CANDELI, H. TEUBNER Internationale Atomreaktorbau GmbH, Bergisch Gladbach, Federal Republi f Germanco y~~

~~Abstract The High Temperature Helium Test Facility (KVK*) which started operatio Augusn ni t 19B reaches 2ha d an operating tim f 1100eo 0 hrs.~~

~~It serve testinn i s g component high-temperaturf so e gas-cooled reactor direcr sfo t heat applicatiod nan steae th mr cyclfo e HTR-Module.~~

~~D-tubW M 0 1 eTesta hea f so t exchanger cyline ,th - drical hot header of the He-heat exchangers, and a secondar ducs ga t t havyho e been successfully completed.~~

Operation wil continuee lb d witW M 0 h1 testa f so helical-tube heat exchanger, a 10 MW steam generator and primary and secondary hot gas duct test objects like coaxiaa bend d san l duct availabie .Th - lity of the test facility is better than 80 %.

Componene Th t Test Facility (KVK) The KVK was built at the Interatom company's site in Bergisch Gladbach in 1980 - 1982. It serve testinn si componente th g higha f so - temperature gas-cooled reacto direcr rfo t heat applicatio e steath mr cyclfo wels a ns la e HTR-Module. tesP PN t e componentth d an K e sponsoresKV ar e Th d* Ministee bth y Economicsr fo r , Small Businesd san Technology of the State,of North Rhine/Westphalia Federae ith n l Republi f Germanyo c .

~~407 The main operating Conditions are (Fig. 1):~~

Thermal power: 10 MW (max. 12.8 MW) Heat transfer medium: Helium Temperature th n i e Primary Loop: C (max° .0 ) 95 100°C 0 Temperature th n i e Secondary Loop) :°C 0 95 C (max° .0 90 System pressure: r (max40ba bar6 4 .) Coolant flow: 3 kg/s kg/s3 (max4. ) . Max. temperature transient: K/mi0 20 n± Max. pressure transient: 5 bar/s

~~Operating data:~~

~~Primary System Secondary System~~

~~Thermal power (maW 1x0) M 12,MW 8 Temperature (maC ° x 0 90 950°C (maC ° x) 0 ) 10095 °C 0 Pressure 40 bar (max 46 bar) 40 bar (max 46 bar) Row rate 3 kg/s 3 kg/s Helium velocity 60 m/s 60 m/s~~

~~FIG. 1. COMPONENT TEST FACILITY (KVK)~~

The test facility consists of a primary and a secondary helium heae looTh t p . gene(Fig2) . - rating sources are a steam-heated preheater, a natura electricaa fires d lga an d l heater. Heat is transferred froprimare mth ye th loo o pt secondar He/He y th looea heapvi t exchanger. The heat sink is a steam generator. Helium is circulate radiay db l blowers.

~~This two-loop test facility allows the installation of reactor test components under their respective operating conditions.~~

408 àas Heater Ellectric Heater Hot Gas Duct Ducs Ga t t Ho Î *f ^ Pressure •-N ^U —— 1>——=* * J — ~* TX1 —————— ( ———— ^ ——— ^> \ Reducing \ {• ^\ x 1 Î Station —» ————— l ————————jp —^ — VJ ————————————— 1-V1 —— HotGasVshie Deaerator Hot Header- — i (Condenier 1 He-Heat . ^L_ Steam Generator- Exchanger" *<££ Feed Water' Tank 1 Î 1 t Î t

~~Feed Water Blower Cooler Blower Pump —— (^ ——— TO —— ^~~

~~Feed Water Preheater~~

~~Storage Tanks Purificatio. n System~~

~~Compressors He-filling~~

~~K DOUBLFIGM . 2 . E LOOP OPERATION WITH He-HEAT EXCHANGER~~

~~Test Components (KVK) e followinTh g main reactor components have been or will be tested under simulated reactor operating conditions (Fig. 3) .~~

He/H- e heat exchange U-tubf ro e design (tests completed) - He/He heat ..exchange helicaf o r l design headet He/Hn a Ho f - ero heat exchanger (tests completed) heliut Ho primare - th m r ducyfo tloop , includin duca g. tn expansioa ben d an d n bellow heliut Ho secondare - th m r ducfo ty loop, including a duct bend and an expansion bellow (tests partially completed) valves ga t s Ho - - Steam generator

409 1984 1985 1986 1987 1988 l l t i i t 1 1 1 KYK-Operatlon t HeadeHo r —— -- •••—-•• • i U M U-Tub0 1 X H e —— mm— --m m i ^^ Helix HX 10 MW m mm •- H— Secondar s DucGa t t Ho y mmi --•mm Secondary Bend —— mm m • i - -- - — " d Du s PrimarGa t Ho y —— •• • • i Primary-Compensator •— - Primary-Bend -——— Hot Gas Valve DH 200 0 s ValvHo70 Ga tH D e —— mm ».._._ Secondary Compensator ^ —— jjj •— •— Fibre-Ceramics-Insulation • Decay Heat Cooler 10 MW — Installatio—— n —.-. Removal ---- Modification ——... Inspection

~~TIM K . OVERVIE3 FIGKV E. E SCHEDULTH F WO E 198 O 198T 8P 4U~~

~~3 Operating Experience and Test Results in KVK~~

3.1 10 MW U-tube Heat Exchanger Fig. 4 shows the test unit. It has been designed by Interatom and Balcke-Dürr and manufactured by Balcke-Dürr. The heat exchanger tubes are bent to a U-shape. Primary helium flows alon tubee gth s outsid countercurrena n ei t arrangement, secondary heliu flowins i m - in g headet sidtubese ho th e positiones e i r.Th d in the upper part of the heat exchanger, resul- rathea tin n i g r short central tube cole .Th d gas header is separated from the support plate and is suspended by springs.

~~The test re-suits are described below.~~

- Steady-state operation was performed at 40 - 100 % with 950 °C on the primary side e secondarth n o C y° side0 an90 d : A comparison of the most important design and test data for the U-tube heat exchanger is

~~410 Secondary A , Secondary Outlet T Inlet~~

~~Support Plate~~

~~Central Duct~~

~~Cold Gas Header~~

~~Outer Annulus~~

~~Pressure Vessel~~

~~Hot Header~~

~~0 perational Data~~

~~Primary Secondary Flow Rate 3,0 kg/s 2,9 kg/s Temperature 9SO/293°C 900/220°C Pressure 39,9 bar r 43,ba 5 Diff Pressure 0,5 bar 1,0 bar Power 10.1MW~~

~~Numbe f tubeo r s 160 Tube Dimension *20 «2,0 Tube Material 24663 INicrofer5S20Co) Vessel Material 1 6366 (WB36) Structure Material 17380 (10CrMo910l 16311 <20MnMoNi55l 15415 (15Mo3)~~

~~Insulation~~

~~Hot Gas Central Duct Primary Inlet Mixing Device~~

~~W HE/HM 0 1 EFIG U-TUB. .4 E HEAT EXCHANGER~~

uppee giveth rn e Fign I i nparth . 5 f .to figure, the temperature profile over the heat exchanger tubes is presented for the 100 % load case. Ther goos ei d agreement, particularly in the hot region of the U-tube bundle. heliua o t me bypasDu s flodesige wth d nan test values differ in the cold region. This also has an influence on the primary outlet

~~411 p- 1000~~

~~LU DC~~

~~111 800 Q.~~

~~î 600~~

~~400~~

~~200 15 20 25 30~~

~~- TUB—> E LENGT] H[m~~

~~TEMPERATURE PROFIL U-TUBE TH F EO HEAT EXCHANGER~~

~~LOAD 100% PRIMARY SIDE SECONDARY SIDE DESIGN TEST DESIGN TEST~~

~~TEMPERATURE [°C]~~

~~• INLET 950 950 220 217 • OUTLET 293 306 900 894~~

~~PRESSURE [bar] 39,9 40,043,6 43,3~~

~~PRESSURE DROP [bar] 0,50 0,401,0 0,86~~

~~MASS FLOW RATE [kg/s] 3,0 2,972,9 2,71~~

~~HELIUM VELOCITY [m/s] 17,6 17,545,6 42,6~~

~~OVERALL HEAT TRANSFER DESIGN TEST~~

~~COEFFICIENT [W/mz K] 446 442~~

~~HEAT TRANSFER [MW] 10,24 9,53~~

~~FIG. 5 COMPARISO. DESIGF NO TESD NU-TUBAN TE DATTH F AEO HELIUM HEAT EXCHANGER~~

~~temperature (306 instead of 293 °C), the secondary mass flow (2.71 insteakg/s9 2. )f o d and the transferred thermal output (9.53 instead of 10.24 MW).~~

The inspection of the U-tube heat exchanger showed the essential components, such as tu- t headebesho ,d insulatio goon an ri de b o t n condition. Only the sheet metal enclosure of the tube sections exhibited signs of leakage

~~412 bene closth d o regionet that .Bu t enclosure is onl expedienn ya testingr tfo wilt .I t lno exist in the full-scale heat exchanger.~~

Non-stead- y state tests were performed suc: as h . Tenfold startup and shutdown of the heat exchange a transien t ra t K/mi1 _ leve+ nf lo over a temperature range of 750 °C. . Cutoff of flow through the IHX at 25 % load over 1.5 h. No measurable convection resulted. Temperature profiles remained stable at a uniformly falling rate. Fift. y thermal cycle tema K/mi7 d t -sa nan peratur. °C e0 ris30 f eo Simulation. : disturbed primary inlet temperature betwee K/mi0 r -5 K/mi0 nfo d +1 n nan a max. temperature difference of 200 °C. - The bearing forces for the load transfer sy- cole th sted bundl e f mheadeo th e d ear r an within the range of the calculated values. - The measured vibrations did not result in any noteworthy loadinheae th t f exchangego - tu r bes. - Successful tests showed leaktightness between the primary and secondary side of the IHX. Ultrasoni- c tubeX inspectios IH swa e th f no performed after a test time of more than 4700 hrs.

~~The following photo (Fig. 6) shows the 10 MW U-tube Intermediate Heat- Exchangefa e th t ra bricator's site.~~

~~Six tube section installede sar bende .Th n sca be seen in detail.~~

~~413 FIG 6 10 MW HEAT EXCHANGER (U-TUBE TYPE) PREPARATION FOR HE-LEAKAGE TEST~~

~~phote takens Th owa , whil heae eth t exchanger was prepareheliue th r m fo dleakag e test.~~

~~3.2 10 MW Helical-tube Heat Exchanger Figur show7 e heae th st exchanger manufactured by Steinmüller/Sulzer.~~

Primary helium enter e innesth r shell space frod bottoe an mtemperaturth a C t ° ma 0 95 f eo i s countercurrenty b coole C ° d3 29 dow o nt - cros helicase th flo t wa l heat exchanger tubes. It then flows back to the primary outlet through the outer annular space, while cooling the pressure shell. Secondary helium enters the helical tubes at a temperature of 220 °C i cole viheaders th aga d exitd ,an s froe th m 1 heat exchangea vi temperatura C t ° ra 0 90 f eo centra e headetht th e ho d lran tube.

~~414 Secondary Outlet r ColHeades dGa r Secondary Inlet~~

~~Support Plate~~

~~Ducs CentraGa tt Ho l~~

~~Outer Annulus~~

~~Pressure Vessel~~

~~Insulation Operational Data~~

~~Primary Secondary Flaw rate 2,95 kg/s 2.85 kg/s Temperature 950/293"C 900/220'C Pressure 39.9 bar 41,9 bar Oiff Pressure r 0.5ba 5 1.65 bar Power 10 MW~~

~~Dimensions and Material~~

~~Number of tubes 117 Dim of tubes *22*2,0 Tube Material 2 4663 I Nicrof er 5520) Vessel Material 16368 (WB36) Structure Material 910o M )r 1738C 0 11 0 HB76 (IncotoySOOHl~~

~~Hot Header~~

~~Mixing Device Primary Inlet~~

~~FIG. 7. 10 MW HE/HE HEAT EXCHANGER IN HELICAL TUBE CONSTRUCTION~~

The next photo (Fig. 8) shows the heat exchanger bundle with 117 helical tubes in the final stage of mounting. This test unit is now installed in the Component Test Facility. Begin- e testh nint f o grun schedules si r fo d October '86.

~~415 , ,~~

~~FIG. 8. 10 MW HEAT EXCHANGER (HELICAL TYPE) TEST BUNDLE WIT HELICA7 H11 L TUBES~~

~~Instrument Penetration~~

~~Water Cooling~~

~~Strain Gauge Insertion~~

~~Test Vessel~~

~~Insulation~~

~~Liner~~

~~Hot Header~~

~~864 Tube Nippel~~

~~Helium - Inlet~~

~~Operational Data~~

~~Temperature 950 °C ' tOOO'C Oiff Pressure <•>• bar~~

~~Dimensions Material~~

~~Tube Nippel .'22-2 24663 (Nicroftr 5520 Col Hot. Header "1020« 100 Wd 3 INicroft2«6 r 55?OCol Test Vessel '2300.41.90 1 5415 H5Mo3l Liner 14876 llncolyaOOH)~~

~~FIG. 9. HOT HEADER CREEP BUCKLING TEST~~

416 3.3 Hot Header Fig. 9 shows a cross section of the hot header as provided for large helium heat exchangers. Heliu flowins i m g aroun outsidee - th din e .Th side of the header had been equipped with a measuring insert for the creep buckling test. The 864 nozzles are closed by caps.

~~In order to perform the fatigue tests, the test arrangemen beed tha n altered (Fig. 10) cape .Th s of the nozzles have been cut off. The helium is now flowing through the nozzles.~~

~~t Helium-Inlet~~

~~Instrument Penetration~~

~~Test Vessel~~

~~Insulation~~

~~Operational Uatu~~

~~Flo» 1 k<|(l >>f '5°°c Presiur (igf3 y 700°b i e ( lemperoture 700*1 • 950 "C Ramp~~

~~Dimension* Holet iol~~

~~lube N,jjptl • 2?. 2 21.66) INnrote' S5/OUI Hot Heade ' 102r 0. 10 0 Wd 2 1.66) INuiolfi S5?l 0(o Test Vesse ' 2)00-1.1.9l 0 l ) 1o 5<.1M S SU liner 1 ".876 dm ol oj 800 HI~~

~~FIG. 10. HOT HEADER FATIGUE TEST~~

~~417 T['C] [barp A ] T [k/min]~~

~~936 43 -0,42 10h I 686 992 (936) CREEP BUCKLING TEST °C 992 43 -0,42 I 742 742 (686) °C~~

~~52 min 950 »- 20 I 950'C FATIGUE TEST 715 ±40 (656 Cycles) 1 950 715'C 20 6~~

~~970 43 0 I CREEP BUCKLING 903 43 •0,21 TEST 903'C~~

~~970'C 970 40 455 h~~

~~FIG. 11. CYLINDRICAL HOT HEADER-MAIN TEST PARAMETERS~~

~~The test programme comprised three test series (Fig. 11).~~

1. The creep-buckling tests on the non-fatigued material of the hot header consisted of two tests with starting temperatureC ° 6 93 f so and 992 °C. The temperature drop was 250 °C in each case, whereby the differential pressure of 43 bar and the temperature transient of 0.42 K/min were identical.

~~orden 2I . fatiguo rt t materiaho e th e th f lo header cycle6 ,65 o t s C wer° n fro0 eru 95 m 715 °C and then back up to 950 °C. The temperature transient amounted to +_ 40 K/min.~~

418 3. Creep-buckling testing of the fatigued ma- t headeho teriae rth consistef lo e on f do test with a temperature drop from 950 °C to 90witC ° 3 transienha f 0.2to 1a K/mi d nan pressure differenc bar3 4 wels f a ,eo s la continuous loading for a period of 455 h at constana a td an temperatur C ° 0 97 f eo differential pressur bar0 4 f .eo

~~Fig. 12 shows a photo of the hot header. You tube th e e canozzlese n there th somd san -f e o~~

~~FIG. 12. HOT HEADER AFTER CREEP BUCKLING TEST~~

~~419 mocouples to measure the temperature distribution. The surface shows a grey protective oxide layer.~~

3.4 Primary Hot Gas Duct Fig. 13 shows the test section of the primary ducs ga t t witho h fibre - insulationli s ga e .Th ner consists of graphite and of CFC. Adjustable supportin lines g ga elementre radiallth x sfi y and axially.

~~8000~~

DetailB Cross-sectioA — nA Axial support element Operational Data sids ega t Ho CoWgassde Pressure vessel AUO, Rowraie: 3 kg/s 3 kg/s Support structure C ' 3 29 C ' 0 Température95 : Insulation Pressure: 40 bar 39 bar Liner Detais m/ 5 l 6, C s m/ 0 2 Vetodly Displacement Radial support element body Dimension Materiad san l Displacement body. » 660x4 1.4876 (Incolcy 800 H) AUO Liner 0 880 x 50/30 u. 10 Graphte/CFC Support structure: «1020x25 15415(15Mo3) Pressure vessel. «1130x25 1.5415(15 Mo 3)

~~FIG. 13. PRIMARY HOT GAS DUCT WITH GRAPHITE LINER AND FIBRE INSULATION~~

ducs primare Testga th t t n undeso ho y r normal operating conditions have been completed. Fig give4 1 . comparisosa f tesno t resultd san design s significandatai t I . t thaheae tth t cole th d o t lossid substans t si eho froe mth - tially lower than calculated (12 KW/m as compared to 36 KW/m).

~~420 970~~

~~960~~

~~950 HOT GAS LINER~~

~~940 P (>• Ni LU o: 930~~

~~CE LU CL LU~~

~~330~~

~~320~~

~~- GRAPHITE- -CFC~~

~~l I A B C C~~

~~DESIGN TEST~~

~~TEMPERATURE [°C] • HOT GAS 950 961 • COLD GAS 293 327~~

~~MASS FLOW RATE [kg/s] 3,0 2,9~~

~~PRESSURE [bar] 40,0 38,0~~

~~HELIUM VELOCITY [m/s] • HOT GAS 19,1 19,6 • COLD GAS 9,1 9,5~~

~~HEAT EXCHANGE PER METER [Kw/m] 36,0 12,0~~

~~TEMPERATURE DROP PER METER [°C/m] 3,0 1,0~~

~~FIG . DUCPRIMAR14 .S GA T T WITYHO H GRAPHITE LINE FIBRD RAN E INSULATION~~

421 Typical temperature profile cole th d f supso - pors tga t tubpressurd ho ean e th et a tubd ean lineleft-han e showe th rar n no de th sid f eo figure. The bandwidth of the temperature distributio includeds ni . Whil maxe eth . band- s linerga e dropt ,i th o t st a C ° 6 widt1 s hi pressure th t a d valueeC an ° 5 dowf s o o nt support tube.

~~Fig. 15 shows the primary hot gas duct as in- foregroune stalleKVKe th e th n se I .n i d u dyo one of the two header of the cold helium.~~

~~. K PRIMARDUCFIG15 S .KV GA TN I T YHO~~

422 5 3. Steam Generator A cross section of the steam generator is shown ie nfollowin Th Fig . 16 . g operational experiences and test results have been gathered: - Fault-free operation for 11000 hrs at high thermal loads and temperatures (up to 900 °C). - No instability on the water steam side in all load ranges. 1 He-Inlet

~~He-Outlet~~

~~Pressure Vessel~~

~~Coded Tube Bundle~~

~~Insulation~~

~~Support Structure~~

~~Operating Data~~

Ha Steam ratw aHo 3.7 kg/s/4,0 kg/s Inlet Temperature 900-0/960-C 150-C TemperaturM Ou e 210 -C 540-C Met Pressure 41 Ober 125 bar OutM Pressure 40,5 bar 125 bar (tower 10,5 MYV Dimensions and Materials

~~Numbe Tubef ro s 12 Tube Length 71 m Heating Surface 66.5m' Tube Oknenelone «25x4/025x26 Tube Material lncotoy800H/15Mo3 WSE36~~

~~.Superheated Steam-Outlet~~

~~FIG. 16. 10 MW STEAM GENERATOR~~

423 - Max. deviation from the average outlet steam temperatur. °C 0 2 H eJ - Good agreement between measured and calculated tube-wall temperatures. Goo- d agreemen overale th f to l coefficients of heat betweetransfe) % 2 n« r calculation and test. - Total pressure drop 12 bar compared to 15 bar calculated. - No inadmissible vibrations of tubes and other structures. Alkalin- e mod operatiof eo compliancn i n e with the VGB Guidelines. - Perfect conditio steae th mf no generator : bundle, insulatio pressurd nan e shels la shown during inspection performed in 1986.

~~Steam Reforme EVA-In i r I Plant A test module of the steam reformer for the PNP-Project with a power of 5 MW was designed by Interatom and Steinmüller and manufactured»; by Steinmüller.~~

~~A longitudinal section throug tese hth t steam reforme wels a rthrougs a l reformeha r tubs i e give Fign i n. .17~~

Primary helium enters the steam reformer at flowt I . 95s °C 0upward e shelth ln s o sidf eo the reformer coole s . i tube °C d d0 san 70 dow o t n Belo e insulatewth d tube shee primare tth - he y liu conveyes i me stea th mo t dgenerator . Having cooled to 300 °C, the helium is returned throug concentrie th h c annulu reace s th bac -o t k tor.

~~424 return tube tube sheet~~

~~recuperator reformer tube~~

~~separation plate guide rube~~

~~guide plate catalyst~~

~~810°C 950T;~~

~~) Pracet(T chambe« s9 r~~

~~(D Product gas chamber~~

~~) Suppor(§ t plate~~

~~FIG. 17. HELIUM HEATED STEAM REFORMER (TEST UNIT)~~

The process gas enters the steam reformer via the proces chambers sga coolt .I e suppor sth t plate before being preheated in a small recuperator whic .integrates i h eacn i d h reformer tube, Hereafte entert ri catalyse sth t regions ,i t heliuho d e man th y b C heate° o abou t 0 p 81 tdu reformed to a synthesis gas with a high content of hydrogen. It leaves the steam reformer at 460 °C.

~~The test bundle consists of 18 reformer tubes e inspectioanon d n tube - materiae tu . Th 3 1 f lo bes is Nicrofer 5520 (2.4663), the remaining 5 tube made sf Incoloar eo H (1.4676) 0 80 y .~~

~~425 STEAW M 5 M . REFORMEFIG18 . R (TEST UNIT) TUBE BUNDL TESR EFO T FACILITY~~

~~The test component (Fig.18) is under testing by RFA, Julien in the EVA-II facility plant.~~

~~Four phase f steaso m reformer tests have been scheduled, the first three ones have been complete Aprin i dwit6 '8 ltestinha g timf eo 3800 hours.~~

- The steam reformer performance under normal operating condition demonstrates swa tesn i d t phases 1 and 2: Heliu. m temperatureC ° 0 95 s betweed an 0 n90 . Helium pressures between 30 and 40 bar . Helium flow between 2 and 3.85 kg/s Methan. e flokg/5 0. ws d betweean 3 0. n . temperature transients between 0.1 and 25 K/min

~~426 hige Th h transients appl disturbeo t y d operation.~~

includ4 Phased an es3 non-stationary operation, e.g. . startup and shutdown . reduced system pressure steam/methane-ratiw lo . o . simulation: tube failure . emergency shutdown behavior leav. e standin statt ho en i g . blocking of one reformer tube

~~Experience with the Materials in the KVK~~

~~In connection with the operation of KVK several tests have been carrie materiae witth t dou f ho l Nicrofer 5520 (2.4663).~~

effece Th f impuritieto heliue th n mi s atmosphere on the corrosion behaviour has been measure y insertindb g material specimens into the KVK loop. Fig. 19 shows surface microstructure and change in carbon content of monitor samples removed after about 4000 hrse Th . metallographic examination reveals only slight internal oxidation; the change in carbon content lies withi scatterbane nth laboratorf o d y test results.

Internal pressure creep rupture test circumn so - ferentially welded tubes (20 mm 0 x 2 mm) of the URKHELIX-head Oan t exchangers have demonstrated the superior strength of the weldment in compari- base th e so o materialt n . Fig show0 2 . s that the tubes failed by developping cracks outside the weld in the base metal region.

~~427 to oo~~

~~0.03 CHANGE C-CONTEIMT [WEIGHT %]~~

~~0.02-~~

~~SCATTERBAND OF LABORATORY 0.01 - TEST RESULTS~~

~~INITIAL CONDITION~~

~~1000 2000 3000 4000 5000 T = 950 °C, p = 36/65 bar, ov = 14/25 MPa TIME (HOURS) t1150/4120/429B= 0h~~

FIG. 19. SURFACE MICROSTRUCTURE AND CHANGE IN CARBON FIG . INTERNA20 . L PRESSURE CREEP TEST WELDEN SO D CONTEN MONITOK KV R RTFO SPECIMENS HELIX-HEAT EXCHANGER TUBES HTR FUEL DEVELOPMENT AND QUALIFICATION- TREATMEN SPENF TO T FUEL

~~. KAISEG R Kernforschungsanlage Julien GmbH, Julien . HACKSTEIK N Nuklear-Chemie und -Metallurgie GmbH, Hanau Federal Republic of Germany~~

Abstract e papeTh r give descriptioa s f desigo n d performancan n e requirements for spherical HTGR fuel elements together with the fabrication technology and the irradiation qualification. Different options for spent fuel treatment are presented and discussed.

~~1. Introduction~~

Mixed uranium/thoriu % uranium 3 oxid9 f mo e enrichment (Hhgh £nriched L[ranium/Thorium fuel) had been the reference fuel for pebble-bed High Temperature Reactors in the Federal Republic of Germany until 1979. Thereafter, non-proliferation aspects and difficulties envisaged with the longterm supply of high enriched a chang uraniuo t d e le mint n uraniua o m fuel wit n initiaa h l enrichmen f aroun o t% (Lo 0 1 nichedwn E d Uranium Fuel) /!/s .A a consequence, emphasi n researci s d developmenan h t wor n fueo k l fabrication and qualification is now placed on this type of fuel which will be used for THTR follow-on reactors. Recover f valuablo y e fuel material y reprocessinb s e th s wa g prime targe n speni t t fue f higlo treatmene h us e th s lona ts a g enriched uranium/thorium fued beeha ln under consideration. Neither economical reason r aspectno s f fueo s l conservation necessitate this treatment, however, when HTRe operatear s d with low enriched uranium in a high burnup modus. Therefore, interim storag n engineerei e d surface storage facilities, followey b d final disposal in a repository is the reference management concep e r neaspen th fo R tfue r n HT ti lfuture .

~~429 . 2 Fuel Element Developmen Qualificatio& t n~~

~~1 Desig2. d Performancan n e Requirements~~

e overalTh l objectiv e Germath R fuef HT o enl development o qualifd prograstilt an s i s ln elemen a ywa m t which minimizes fission product release under normal, transient and accident conditionR typel HT al f e use o sr b d whic fo dan n s ca h applications /2/. The design of the FRG fuel element is shown in Figure 1. Abou 0 1 tcoate d fuel particle e dispersear s a graphit n i d e matrix e fuele.Th d zon s surroundei e a fuel-fre y b d e shell compose e samth ef o graphitd e material e overalTh . l diametef o r m thicthc e 5 km witelemen c 0. fuel-fre a 6 h s i t e shelle Th . coated so-typ i diameteparticlen Tr ur d contaif o an e0 e 50 r ar sa n oxide fuel kernel. Tri so refers to a four-layer coating with a silicon carbide (SiC) layer sandwiched betwee o high-densittw n y pyrocarbon (PyC) layers.

~~• MATRIX~~

~~KERNEL~~

~~BUFFER LAYER~~

~~INNE LAYEC RPy R~~

~~SiC LAYER~~

~~OUTER PyC LAYER~~

~~FUEL-FREE SHELL~~

~~FUELED ZONE FIG. 1. HTR fuel element.~~

Tabl 1 summarizee e desigth s e presenn th dat G f o aFR t reference fuel and the operating requirements. To keep already developed fabrication processe s constana s s possiblea t e ,th particle design parameters of the LEU-fuel have been chosen to be identical the design data of the former HEU fuel.

~~430 TABL DESIG. 1 E N DATR (Th,U)FO A 0 „ TRISTRISU0 D O AN OFUE L ELEMENTS~~

~~Fuel Type Design Parameter HEU LEU~~

~~Coated Particles Kernel Composition (Th,U)O2 UOj Kernel Diameter (im 500 500 Coating Layer Thickness um 90/40/35/35 90/40/35/35 Coating Layer Sequence Buffer/PyC/SiC/PyC Bufter/PyC/SiC/PyC~~

~~Fuel Element Heavy Metal Loading g/Element 11 8-12 U 235-Ennchment 93 % 7-13 % No Particles per Element 19000 13,000-20,000 Volume Loadin f Particlego s 13 % 10-15 %~~

~~Operating Requirements Mean Operating Time a 1100-1500 700 Max. Burnup Gw0 1 MeV) 10" m ' 45 33 Max. Fuel Temperature °C 1020 1030 Max Power/Element kW 2.7 4 1~~

The Tri so coating was introduced to fulfil all requirements defined by the different reactor projects, especially the stringent requirements for gas-turbine and process-heat applications. Because elevateth f o e d operating temperaturef o s

these advanced HTRs e silicoth s n carbide layer only provides sufficient retention of the more volatile solid fission products cesium and silver. To guarantuee a clean primary circuit for easy maintainance, the following specifications have been used as a development goal : o Defect particle fraction due to fabrication 60 x 10-6 o Defect particle " fractio 10 o operatio t x e 0 du n20 n ,-6 This toleratetotaa f f 1,000,00o o l t ou s 0 coated particles t morno e tha 0 individual6 n o havt s e fabrication induced defects andn additioni , t morno , e tha0 additiona20 n l particle o fait s l during reactor operation.

~~2 Fabricatio2. n Technology~~

The fabrication process for spherical HTR fuel elements consists of the following steps /3-7/: o kernel formatio l precipitatioge y b n n o chemical vapor depositio e coatinth f o ng layers o preparation of resinated graphitic matrix powder o cold molding of the fuel element spheres

431 Kernel formation is achieved by a gel precipitation process as outlined in Figure 2. After formation of an uranium nitrate broth, spherical droplet e producear s a vibrationa y b d l dropping technique. The droplets are aged to improve the internal structure and then washed to remove the nitrate. After drying and calcining, a special U0„ related step - reduction of the calcined kernels to UO« - is applied. Kernel fabrication is completed by sintering to produce highly dense material.

UOzlNOjtj k PVA • to BROTH OFF GAS ADDITIVES ————————————• . FORMATION i NH3 ————————————» • DROPPING -»- MH —* NHj • u Mi-n . i OF KERNELS 3 ABSORPTION

~~AGING~~

~~. NH«NO_» H « NH4 OH ————————————» . 3 20 WASHING ISOPROPANO ——O H2 — - f R LO .». ISOPROPAN(DL RECYCLING + ——— 1 ————————r DISTILLING~~

~~DRYING -*• ISOPROPANOL 1 1 WASTE h- C02~~

~~CALCINING *• MH3~~

~~»• MZ0 REDUCTION Mi ————————————• - TO UOi -^ MjO l «• HjO HZ ———————————• * SINTERING UOj KERNELS *• H»~~

~~„ kerneU0 FIG. 2 l. formation.~~

Coatine kernelth f o g s (Fig s performei .) 3 fluidizen i d d beds by pyrolysis of hydrocarbons for PyC and of si lane dérivâte r SiCfo s . Becaus f criticalito e y restrictionse ,th maximum coating batc f heavho sizg k ylimites i emeta0 1 o lt d oxide. A coater with this capacity, having a coating tube of 400-mm inner diameter s beeha , n develope d sucessfullan d y tested. In the manufacturing process (Figure 4) phenol and hexamethylene-tetramin e firsar e t mixe t elevatea d d temperatures with graphite powder. Durin e warth g m mixing process, resin binder is formed by chemical reaction of the components. The resulting materia s thei l n groun a fin o et d powder .A portio n of the resinated powder is used to overcoat the coated particles; another portion is premolded together with these

~~overcoated particles to for1 m the fueled zone of the fuel elements. The remaining material is taken for the final molding of the fuel element giving the element its fuel-free shell.~~

~~432 KERNEL BATCH~~

~~BUFFER LAYER (POROUS~~

~~INNER PyC LAYER (HIGHLY DENSE) JL SiC LAYER~~

~~OUTE LAYEC RPy R (HIGHLY DENSE)~~

~~SIEVIND GAN CLASSIFYING~~

~~COATED PARTICLES~~

~~FIG. 3. Coating of HTR fuel kernels,~~

RAW MATERIALS COATED PARTICLES *

~~1 MIXING WITH GRINDING OVERCOATING RESINATED POWDER AND PREMOLOING,~~

~~HEAT LATHING FINAL PREPARATION OF THE TREATMENT MOLDING FUEL ELEMENT SHELL~~

~~FIG. Fabricatio4 . f sphericao n l fuel elements.~~

Final steps in the process are: lathing the elements to obtain the appropriate diameter and heat treatment for binder coating and removal of impurities. The crucial point in fuel fabrication is to reduce particle defects to the lowest possible level, that means at least below the design value of 60 x 10 . To reach that goal, two mechanisms which might cause particle defectse b hav o t e overcome: . o during the molding process, adjacent particles can mechanically interact, inducing local peak stresses that

433 might crack the coating layers. This effect can be avoided by overcoating the particles with layers of resinated graphite powde rs spacers a whic t ac e hbes overcoatinn Th .i t g qualities have been obtained by classifying the overcoated particles using vibrating tables. o extremely odd shaped particles - whether they are overcoated or not - cannot withstand the pressure applied during molding o excludT . e this e coate,th d particles have also t o e classifieb y vibratiob d n separation before overcoating. Figure 5 shows how the product quality is influenced by the classification procedures. In 1977/1978 statistically no fuel element as fabricated was fully free of defect particles, and many elements («30 %) had three or more particles with a cracked coating. Once the technique of precise classification of overcoated particles was established, the defect rates reduced significantly. On a developmental as well as on a production scale, more than 50 % of the fuel elements were found free of particle defects, and only a very low percentage (« 5 %) showed thre r moro e e particle defect r fuepe sl element. Introduction of the additional classification procedure for coated particles resulted in a significant further improvement of the product quality e fue: th morl f o elemente % tha 7 9 n s have been fref o e any faile e particld on particlese remainin% th 3 n eo i t d 2 g an , defect onlr elemenpe y t coul e foundb d o date.T e defec,th t particl whic, equivalenis h10" e x fractio1 1.5 to t is n defective particle in a total of 40 spheres. This positive

~~100~~

~~STATUT SA STATUS AFTER PRECISION STATUS AFTER BEGINNINF GO CLASSIFICATION OF ADDITIONAL DEVELOPMENT OVEROOATEO PARTICLES PRECISION CLASSI- FICATION OF COATED PARTICLES~~

~~X 50~~

~~01234 0123 0 t 2 0 1 FUEL ELEMENT WITH 0.1.2... SiC DEFECTS 1977-1978 1980-1981 1981 1982-1983 DEVELOPMENT DEVELOPMENT PRODUCTION DEVELOPMENT AND PRODUCTION~~

~~. PercentagFIG5 . C defecSi f o te number n fuei s l elements.~~

434 difference between desig d actuaan n l measured values e latteth , r of which correspond to pilot-scale manufacturing, is a good basis of guarantee that the design goal can be met in a large-scale productio o /8/to n .

~~2.3 Irradiation Qualification~~

Irradiation testing of fuel samples and single fuel spheres was and is performed in Material Test Reactors, mass testing of thousands of fuel elements in the 15 MJ/s experimental HTR power plant AVR. Important performance related parameters appliee th n i d simulation of normal operating conditions have been: o irradiation temperature: 800"C-1200"C o burnup: 8 - 14 % fima o accumulated fluence of fast neutron: 1-8 x 102 5 n/m2 The operating behavior of LEU fuel has been investigated in a comprehensive basic irradiation program, including tests with the following main objectives: o Reference test: irradiation of reference fuel elements under conditions envelopin e demande differenth gth f o sR HT t project n temperatureo s , fast neutron fluence d burnuan , p o Determinatio particlof n e defects: testin fueof g l under conditions exceeding the demands of the HTR projects with referenc o fast e t fluenc d burnuan e o investigatt p e design limits o Investigation of burnup influence: irradiation of fuel in thermal test reactors witfasw lo ht neutron fluxeo t s separate burnup-controlled effects from neutron-induced effects o Investigation of solid fission product release: generation of input data for release calculations

e resultTh f theso s e experiments, summarize Tabln i d , 2 e demonstrat e excellenth e t irradiation performance. Neithee th r in-pile measure e resultB valueth R/ dr f posno o s t irradiation examination work indicat y particlan e e breakage. For mass testing, more than 40,00 U fue0LE l elements have bee ne meantiminserteth n i d irradiate R an ed AV o intt e p th ou d

~~435 TABL . FISSIO2 E S RELEAS GA NO U FUE U EL LE RAT F O E~~

Temperature Fast FUience Burnup R/B* Experiment ("0 (I021 cm'2) (% fima) <«""Kr) Reference test 1200 4.0 8.5 3x lu' 7 1000 6.0 10.1 2x 10-' Determinatio r particlno e defect rate function 1200 8.0 11.6 9 x UP1 SiC(3n 5ur ) 1000 8.0 11.6 8 x W* Determination of particle defect rate function 1000 8.0 11.6 8x It)'* (50 Mm SiC) Investigation of burnup influences 1200 <1 12.0 2x K)'1 Investigation of solid fission product release 1200 1000 1 t completeTesye t t no d 800 ."End-of-lifc.

~~% fimaburnup 0 1 . f Behavioo s r under reactor operating conditions is also excellent; to date no fuel element damage was found.~~

~~2.4 Core Heatup Accident Simulation~~

In small modular HTRs the maximum fuel temperatures will not exceed 1600"C even under the severe accident conditions of core heatup and simultaneous depressurization /9, 10/. In the HTR 500 the design basis accident is defined by some los f coolano s t while forced convectio s stili n l maintainey b d at least one loop removing the afterheat. Maximum fuel temperature is kept below 1250*C. The hypothetical accident sequenc n unrestrictes givea i e y b n d core heatue th n i p depressurized reactor in combination with a failure of all active cooling systems. This sequence lead o maximut s m temperature e pebbl th e fue d corn th i be lf e 2350" o f eo s C /11,12/. t thesA e high temperatures, fission product n leavca e s th e fuel elements. Two failure mechanisms are prominent: o thermal decomposition of the SiC layer, and o SiC corrosion resulting from the interaction of fission products, mainl , wite silicoPd yth h n carbide layer In addition, diffusion processes in the coating layers can increasingly contribute to fission product release. Core heat-up simulation tests with irradiated fuel elements have demonstrate e excellenth d t performanco t R fue p HT u l f o e e releasy nuclideTh ke 180: f *C o 0es negligibl i o st p u r fo e

436 hundreds of hours at 1600°C and small for dozens of hours at 1800°C (Fig. 6). Ceramographic sections through coated particles show no damage to fuel coatings. C thermaSi l decompositio s settin i nt temperature a n i g s well beyond 2000°C. This leads to near 100 % cesium release in temperature ramps to 2500°C.

~~1800 Cl~~

~~«•-01 AO nom~~

~~Ag-liOn~~

~~7 C13 3 FRACTIONAL RELEASE~~

~~t»-OB- • Cfl 137~~

~~«•-07- ...f.. ieoo c Kr es Kr 85 7 ...£}13 .a C 160 C 0 o n B A 160 c 0 Kr B5 7 13 s C IBO OC O I1 O A IBO OC s a - ieoKI oc to~~

FIG. Accumulate6 . d releas f fissioo e n product t 180a sd 0an 1600°C from spherical fuel elements witU TRISLE h O particles.*

*The contributions from contaminatio e fueth ln i nelemen d froan t m release during irradiation have been subtracted from measurement data.

~~. 3 Spent Fuel Treatment~~

~~3.1 Spent Fuel Treatment Options~~

For the management of spent HTR fuel 2 principal options exist: o reprocessing or o final disposal e FederaIth n l Republi f Germano c e maximuth y m amounf o t spent fuel to be expected by the year 2000 from the presently

437 existing HTR plants AVR and THTR is estimated to approach 3.5 million fuel spheres or 25 metric tons of uranium/thorium oxide. For this limited amount, final disposal is definitely planned. e follow-uth r Fo p plants, which wil e operateb l d witw lo h enriched uranium in a high burnup modus, technical and economic argument e samsth ecalr treatmentfo l .

~~3.2 Interim Storage~~

o storT e fuel discharge o preparet d an n R ,o d AV froe th m a R&D basis, for the interim storage of spent HTR fuel, KFA was and stil s operatini l o engineeretw g d storage systemf o s different designs: a vault-type storage facility and transportation/storage cask prototypes e vault-typ.Th e storage facility /15/, consists of a stainless steel rack with 36 storage positions. Each position can be loaded with 2 storage canisters containing about 1,000 fuel elements each. This facility has been at full capacity since April 1984. The total activity of stored fuel amounts to about 600,000 Curies, including about 10,000 Curies of Krypton-85 and about 2,000 Curies of Tritium. No release of radioactivity has been monitore e maximuth o datd t d an me surface temperature th f o e sealed canisters has been only 10"C above the ambient air inlet temperature /16/. Cast iron casks have also been use Germann i d r somfo y e years as transportation/storage systems for spent nuclear fuel. An experimental program performed by KFA with prototypes of HTR casks (Fig ) demonstrated.7 , impressively e safetth , y features of interim storage of spent HTR fuel in those systems (Table 3).

~~3.3 Final Storage~~

Spen R fues beeHT tha l n include e Physikaliscth y b d h Technische Bundesanstalt e designateth , d operato e futurth f o er Federal repository, in the group of waste with medium heat production which 0 wil metee dispose30 b l n ri deef o d p boreholes. The development of the general storage techniques and the testing of necessary below-ground components like transport system, charging machine, borehole plug etc. is performed in

~~438 FIG. 7. HTR transportation/storage casks prototypes.~~

~~TABL . CHARACTERISTIC3 E D OPERATIONASAN L DATF O A HTR-TRANSPORTATION/STORAGE CASKS PROTOTYPES~~

~~Ch»r«ct«rl»tlc» Material Cast Iron Dimensions 2 6 m X 1 3 m Dia Wall thickness -03 m~~

Cask weight ~20Mt Capacity THT1 R spent fuel canister ro 2 AVR spent fuel canisters (200) 0FE Cask loading tor OEMO-test 2 AVR spent fuel canisters Dat f loadeao d material decay time -1 YR tola! activity -4200i 0C Kr-85 activity - 700 Ci 3 activitH- y 95 Ci decay heat -180 Watts

Operational Data Cask leakage S10E-7mbas 1/ r Canister leakage mba4 s E- 1/ r0 » S Cask surtace temperature 1 *C above ambient temperature max y-dose rate (contact) S12mR/hr max n-dose rate 04 n/cm E-2 s Activity release ove monthB r s inside cask Kr-85 5* 5' 10E-7Ci H-3 S3 x 10 E-8 Ci outside cask 1 zero-release"

439 connection with the R&D work related to the disposal of specific heat generating wastes resulting from LWR fuel reprocessing. Therefore e e emphasi th ,th R wor s been HT d stilo ha k an n f s o si l followin majo4 g r items: o experimental evaluation of HTR fuel performance under normal and upset operational condition e repositorth f o s y o developmen d testinan t f packagino g g concepts o safety analysi e long-ters th wor n o k m R fuesafetHT l f o y disposa n sali l z formations o in situ demonstration of ultimate HTR fuel disposal in an experimental salt mine facility. Available results on the release of gaseous fission products /23/ and on the Teachability of solid fission products by salt brines /24/ indicate that the multiple barrier concept e coateth f do particles, develope r fissiofo d n product retention in reactor operation, also ideally suits HTR fuel for final disposal t onle gaseouNo .ar y d solian s d fission products safely retained t chemica,bu l attack froe outsidth m s preventei e d too. The full demonstration program is, however, time consuming and it will take some additional years of R&D work until the HTR fuel will finall e qualifieb y a produc s a d te safel b whic n ca yh stored in a salt mine forever. The results of a planned in situ pilot experimen e ASSth En i tsal t e frammineth f whicn o ei , h a limited amount of AVR fuels will be stored for 5 years in monitored borehole sf considerabl o , wil (Fige 8) b l. e importance in this qualification process /19/.

~~cfissdver fuedadoingR l elementHT , s precursory lest sludge structural components ; i •**>*•£ à——.r^7~~

~~t air-conditioned measuring conlainer 4 cabinet rack 19* 2 entrance lock 5 borehole slide- storag6 e rack 3 crane ( FIG. 8. Schematic view of the MAW test drift EV (retrievable disposal test).~~

~~440 4. Summary and Conclusions~~

The state-of-art in HTR fuel cycle work in the Federal Republi f Germane o summarizecb n ca y s followsa d : o The development of processes and equipment for fuel fabrication has progressed to an advanced stage. Basic irradiation experiments with low enriched uranium fuel have been sucessfully completed; final tests simulating the operating condition f commerciao s n lpreparationi HTRe ar s . High temperature heating tests with irradiated material indicat n excellena e t behavio e fueth l f o rals o under core heatup accident conditions up to 1600/1800°C. e technologTh o f interio y m surface storag f speno eR fueHT tl is developed. Taking advantage of the experience with the storage facilitie n operatioi s d usinan n g additional data resulting from ongoin D workR& g , interim storage facilities r upcominfo e gdesigneb HTRn ca s d properly e developmenTh . t of techniques for final storage of spent HTR fuel in salt n progresdomei s i sd wil e demonstratean sb l df o unti d en l 1992.

~~References~~

~~1. D.F.Leushacke and G.Kaiser, "HTR fuel cycle program in the FRG" (1979)_ Trans31 C 0 ,EN 19 .~~

~~. 2 W.Heit, H.Huschka, W.Rin G.Kaiserd an d , "Statuf o s Qualification of High-Temperature Reactor Fuel Element Spheres", Nucl. Tech. 69 (1985), 44~~

~~3. M.Kadner and J.Baier, "Ober die Herstellung von Brennstoffkernen für Hochtemperaturreaktor-Brennelemente" Kerntechnik L8 (1976), 413~~

~~4. H.Huschka and P.Vygen, "Coated Fuel Particles: Requirements and Status of Fabrication Technology", Nucl. Techn. ^5 (1977), 238~~

~~441 5. T.D.Gulden and H.Nickel, "Coated Particle Fuel", Nucl.Techn. 5 (1977)^ 5 20 ,~~

~~6. R.E.Schulze, H.A. Schulze and W.Rind, "Graphite Matrix Materials for Spherical HTR Fuel Elements", JUl-Spez. 167, Kernforschungsanlage Julien (July 1982)~~

~~7. K.G.Hackstein, "Herstellung von Brennelementen für den THTR d AVR-Reaktor"un , Atomwirtschaf (1971)6 _1 t 5 24 ,~~

~~. 8 K.G. Hackstein, W.Heit, W.Theymann, G.Kaiser "Stan r Brende d - nelementtechnologie für HTR Kugelhaufenreaktoren", Atomkerne- nergie-Kerntechnik 47 (1985), 163~~

~~9. J.J.Wolters, W.Kroger et al. "Zum Störfallverhalten des HTR-Modul; eine Trendanalyse" ReporA KF , t Jül-Spez-260, Juni 1984~~

~~10. J.Mertens, "Sicherheitstechnische Untersuchungen zum Störfallverhalten des HTR-Modul", KFA Report Jül-Spez. 335, November 1985~~

~~. J.W.Kroger11 , J.Wolter t al.e s , "Zum Stb'rfallverhalteR HT s nde 500; eine Trendanalyse" ReporA KF , t JUl-Spez-220, September 1983~~

~~12. R.Nabbi, "Sicherheitstechnische Untersuchungen zum Störfallverhalten des HTR-500", KFA Report Jül-Spez-240, Januar 1984~~

~~. W.Schenk13 , D.Pitzer, H.Nabielek, "Spaltproduktfreisetzungsverlau Kugelbrennelementen vo f i be n Störfalltemperaturen", JOL-2091 (1986)~~

~~14. W.Schenk, "Störfallsimulation an bestrahlten Kugelbrennelemente i Temperaturebe n n 140 s 2500*C"0bi A KF , Report Jül-1883, December 1983~~

442 15. U. Brinkmann, R. Duwe, R. Engel stätter, S. Storch, "Entsorgun s AVR-Versuchskraftwerkede g d Untersuchungeun s n an abgebrannten Brennelementen für trockene Zwischenlagerung", JTK 1980, Tagungsbericht, ISSN 0173-0924

~~16. R. Duwe, H. Müller, "Lagerverhalten abgebrannter HTR-Brennelement n Transporti e d Lagerbehälterun - s au n Sphäroguß", Jül-Spez-254, 1984~~

~~17. R. Duwe, U. Brinkmann, Freisetzung gasförmiger Nuklide aus lagernden HTR-Brennelementen" K 1985JT ,, Tagungsbericht, ISSN 0720-9207~~

~~. RölligK . . Brinkmann18 U , . DuweR , . Rind,W , "Auslaugunn vo g abgebrannten HTR-Brennelementen unter Endlagerbedingungen", JTK 1985, Tagungsbericht, ISSN 0720-9207~~

. BarnertE . 19 , P.H. Brücher . Niephaus,D , "R&D Wor n Geologio k e Disposa f Dissolveo l r Sludges, Cladding d Spenan s t HTGR Fuel Elements in the Federal Republik of Germany", Int. Symposium on the Siting, Design and Construction of Underground Repositorie r Radioactivfo s e Wastes, Hannover, 7 FRG3- , March 1986

~~443 CHARACTERISTICS OF HTGR SPHERICAL FUEL ELEMENTS~~

~~A.S. CHERNIKOV, L.N. PERMYAKOV, L.I. MIKHAJLICHENKO, S.D. KURBAKOV State Committe Utilizatioe th n eo n of Atomic Energy, Moscow, Union of Soviet Socialist Republics~~

~~Abstract~~

The paper presents the investigation results for the basic characteristic sphericaf so l fuel element components: fuel microspheres (dimension, shape, density, structure, strength), pyro- carton d silico,an n carbide coatings (thickness, density, macrod -an microstructure, crystalline structure, thermal stability, gaseous fission product release, radiation stability etc) and matrix graphite (density, effective thermal conductivity, linear thermal expansion coefficient, elasticity modulus, bending strength). manufacturinn I g fuel microspheres sol-gel proces d slip-castinan s g methods were used; coating depositio s accomplishenwa n fluii d - dized bed, high-density pyrocarbon coatings being deposited from

~~CH.-A C^EL-Arr ro C coating3i ; fros- m CH^-SiCl^-H^A CH-jSiClr ro 3 ———— - Hp - Ar mixtures; matrix graphite was prepared on the basi bulf o s k graphite MPG-6d fillersan G P . 0 ,3~~

~~1. Introduction~~

e concepe pebble-beTh th f o t d cor high-temperaturf o e e helium- cooled reactors (HTGR) with spherical fuel elements is being deve- USSe VGR-5r lopeth fo R VG-40d n i d0an 0 plants /1/. The current effort e directesar improvemeno dt d stabilian t - maizatioe th n f no characteristic e sphericath f so l fuel element (FE) components: fuel microspheres (FM), coated particles(CPd )an matrix graphite (MG).

445 FM producee (50ar 0 ) jt« .m d from uranium dioxide; fission products (FP) are retained within CP by the multi-layer barrier coatin pyrocarbof o g n (PyCd silicoan ) n carbide (3iC) disperP ;C - sed over the graphite matrix constitute a spherical fuel element of HTGR. e realizeTh manufacturinE F d , processeCP , FM basee f o sgar n o d utilization of various ver/aions. FM - sol-gel process and slip casting method; CP - deposition of so-called high-temperature and low temperatur C coatingePy s coatingC LTC)d (HT Si wels an C a , s la s from SiCl.-CH.-Hp-Ar and CH-SiCl-,-EU-Ar mixtures; FE - matrix graphite based on bulk carbon graphite materials, of various types (e.g. 30 PG and MPG-6 type graphites). Compariso maie th n f ncharacteristico E componentF f so s (FM, CP and MG) realized by the above versions is the aim of the present work.

2. Fuel microspheres 2.1. Main requirements In selection of FM spheroidization process various physical and chemical methods were considered and tested, including: variou- s version so-callef so aglomeratioy dr d n from finely dispersed powders; - slip casting methods; - sol-gel method. Compariso maie th n f ngeometricao d structuralan l characteristics of FM manufactured using these methods as well as analysis of various spheroidization versions showed that the most acceptable of sol-gee th thed drose lan mar s spheroidization methods /2/. Both methods allow FM of the required quality, meeting increasing requirements to HTGR fuel to be obtained.

446 As shown by the literature analysis and investigations FM quality is determined by the following main characteristics: - density (apparent and pycnometric ) ; - average dimension; -non-sphericity factor; - crushing force; - structural characteristics (grai d pornan e sLzes,pore distri bution ove particl, rFM e surface relief, etc); phas- chemicad ean l composition. Amonspecifie th g c requirement radiatioo st n performance

of HTGR spherical fuel elements the most imprortant one is that P releasF then i e m woul t exceedno d 10"- (R/B)3 , whic equis hi - valent to one damaged CP per ^10 particles. These requirements onle whe t b startine yme n nth ca g product ,highl s i.ei . y.FM homogeneous. Therefore, along with ensuring specified average values of the FM basic characteristics a high degree of their homogeneity is needed both within one batch and between FM batches.

~~2.2. Spheroidization methods~~

The sol-gel method of FM spheroidization from uranium dioxide is based on the sol-gel process similar to the"H-process " of KFA reported in /3/. The working solution contains uranium, urea and o urotropin relation i e densita n s 1:2:1. ha 1.6f o yd 55an g/cm. Spherical particles produced by internal gelation in petroleum jell washee 90-9t ar ya CCIn C 5i d° d ammoni an . a solution, then subjected to azeotropic drying in CCI, and calcinated in argon - 4.0% hydrogen mixture atmosphere.

Another method of FM production is based on spheroidization of plastic slip which is prepared of finely dispersed uranium dioxide powder and binder. At 70-75 °C the slip mass is highly fluid, it is injected drop - by - drop into glycerine, where it solidifies with formatio homogeneouf no s spherical particles.

447 After spheroidization particles are washed from glycerine and the bindee nth s distilleri themf o t «ou d Particles produced using both methods are sintered in argon at 1500-1700 °C, sieved till the desired fraction is obtained and classified by their shape.

~~2.3 characteristicM .F s~~

~~sho2 e curvedistributioM PigsF d wth r an sfo .1 theiy nb r sizespresen4 figs d d M curvee ,distributioan F an th t,r 3 sfo n~~

~~by non-sphericity factor particler (Dfo m __/) D^ s manufactured slie bth yp spheroidizatio d sol-genan l methods, respectively.~~

~~3.6 -~~

~~2.4 -~~

~~ü a~~

~~5~8.Ö diameter,ju.m~~

~~Fig. 1. Distribution of U02 fuel microspheres manufactured slip casting method by sizes: - abatc h 9789; b - batch 982?; c - batch 9858~~

~~448 7.3~~

~~3.6~~

~~50 l~~

~~2.4 l~~

~~25 J n 0) so* tfc. o o 1.2 4 0) go1 «H 1.00 1.02 1.04 1.06 1.08 1.10 1.12 1.14 non-sphericity factor •H •P~~

~~440 460 480 0 56 0 5254 0 580 diameterm ,u.~~

Fig Distributio. 2 . fue- UO l f nmicrosphereo s manufactured Fig Distributio. 3 . IK>f no 2 fuel microspheres manufactured by sol-gel method by sezes; by slip casting metho sizesy db : a - batch 9792; a - batch 985ß; l batc- b h 9726; b - batch 9027; batc- c h 9857 c - batch 9789 80

~~100 » • . . * •' 90 . .__.. //-- *.i • . -•*«" .^ b ,.______V, o — / .-•~~

~~50 a - 70 - • 4 =cL1 £ • * «~~

~~0} . + - rH •~~

~~•* * 25 30 . . • * o « - . c 1 Ü 4 " . 4>~~

ïig. 4. Distribution of UO^ fuel oicrospheres . Cumulativfi5 . e distributio 2 fueU0 l f microsphureo n s (sol-gel method) by non-sphericity factor: by porosity : a - batch 9726; a - batch of sol-gel-manufactured microsphures; b - batch 9857; - bbatc f slio h p casting c - batch 9792 Sprea porosityn i d , measure quantitative th y db e metallography method usin e imaggth e analyzer s characterize,i e curveth y sdb of FM distribution by porosity presented in fig 5. It follows from comparison of the curves that the geometrical characteristic sphericaf so l particles manufacturee th y db two methods are good from the homogeneity viewpoint. FM manufactured by the sol-gel method have somewhat better homogeneity in the g«ometrical characteristics and FM produced by slip- apheroidization method are more homogeneous in porosity. e typicaTh l microstructur M prepareF f o e boty b d h methods i s . 6 g showfi n ni •BBjj^HPP^^^^!~r^lHjMjHBB

~~Pig . Kicroatructur6 . f UOo e g fuel miorospheres manufactured by slip casting (left) and sol-gel (right) methods: a, t> ) nOO; c, d - 22*0~~

~~4SI Coate. 3 d particles 3.1« Manufacturing methods~~

desigP C s beee nha Th n develope basie th f calculatioso n o d n taking into account operation conditions /4/ and implies, as mentioned abovehighlf o e ,yua dens C coating ePy typeso tw f s:o HTC or LTC, as well as SiC layer. The requirements to the density geometricad an l characteristic protectiof so n layer coatingsf so , we nearle reportear , y 6/ identicad , earlie/5 appli r n lfo i r - cation of any PyC types. In deposition of high-density LTC layer on buffer PyC layer, the medium density (1..4-1.6 g/cnr) PyC layer in the CP design is absent. Depositio protectiof no n s coatingaccomplishei P C n so n i d the apparatus of pseadofluidized bed with a conic gas-distributing device. Tabl present1 e s protection coating deposition conditiond san list reagente sth s used.

~~TABLE 1. PROTECTION COATING DEPOSITION CONDITIONS~~

~~Tempera- Coating layer SM^ maturÄ* e ture, kPa °C~~

Buffer (PyC-1) Cr 2A H- 2 40-60 1500-1550 Dense isotropic layers: - HTC r A - 4 CH 10-20 1900-2000 - LTC C3H6-Ar 15-31250-1400 0 SiCl4 CH O 1 r*tJ TT A -v* -1/"\*_OC oiuj.4-o.n4-tt2-Ar iru<-^.p 1450-1550 CH, SiC 1.0+2.0 = 1. 01-2.0 2.5*3.5 1500-1600

452 Rejectio witP C hf n o damage d coatings implies their treatment in the acid and subsequent distribution by their densities. specifiee th f o P dC shap d sizthee ean e ar n separated froe mth batch using appropriate classifier.

~~3.2 coatinP .C g characteristics~~

~~The typica wels la wit s macrostructurla C LT h d an C HT f eo microatructure SiTh C . 7 laye g shows fi ri n high-densitf ni o e y LTCd an ,C founHT y oxidatiodb glow-discharge th n ni e plasm/ a/?~~

~~Pig Uacxostructur. .7 and) (a . C I/E) coateHT C(6 f o e d particle sections, Ï70~~

453 an dC laye Si tha rf o t afte r electrochemical picklin finea s -gha crystalline structure with equiaxial grain (fig. .8) The PyC cristalline structure was determined in all cases graphite-typs a hexagonae on e witt lbu lono h n g range order). The result X-raf so neutrod yan n diffraction investigation- in s dicate that PyC grains consist of several crystallites (0.002- 0.03 Mm in size), whose number and orientation determine the grain shape and size. SiC coatings obtained using the above mentioned mixtures had a cubic modification with the lattice parameter a=0.4360-0.00 hav e t observeW eno . 1nm y essentia dan l effect s mixtur e levega e typ th o f th ff otheo e n lo o e r propertief so high-densit 3.2> Yl 0 coatingsg/c( C y Si m) .

~~FigMicrostructur. 8 . high-densitf o e y coating) (a C sET 12000 £100(ft) C ;(o LT ) 0C 2200Si d 0an~~

454 The difference between LTC and HTC is accounted for by differences in their structure dispersity. The sizes of LTC crystallites vre have obtained are essentially (by 5-10 times) smaller than thos HTCf o e , while interlaye re longe distancear C rLT n si than thos HTCf o e .

~~3.3. Results of CP teats~~

~~The characteristics of typical CP batches with HTC and HTC as well as those with 3iC coatings are listed in table 2.~~

~~TABLE 2. CHARACTERISTICS OF COATED PARTICLES WITH LOW- TEMPERATURE (LTC HIGH-TEMPERATURD )AN E (HTC) COATINGS~~

~~Coated partic- Coated par- Parameters leo witC hLT ticle« with HTC~~

~~Fuel microspher: e diameter ,j«. m 490 500 sphericity factor 1.03 1.02 density, g/cnr 10.8 10.5 oxygen coefficientU ,0/ 2.0003 2.0020~~

Coated particle . coating thicknessm u ,j PyC-1 95 90 PyC-2 67 35 yC-3 - 40 SiC 50 53 PyC-5 60 50 coating density, g/cnr PyC-1 1.0 1.0 PyC-2 1.85 1.55 PyC-3 - 1.80 SiC 3.2 3.2 PyC-5 ' 1.82 1.8 Lay«r contamination: surface, xlO 13 Ci/cm'3 1.0 1.0 bulk, g U235/g coating, x105 1.0 3 crushing force CP, kg mean 7.7 6.6 mai/iain 10.0/5.2 8.0/4.2 Leakage at the atage of "weak" irradiation6 10 ,X 1.6 2.0

*The uranium dioxide FM manufactured by slip casting methods (CP with LTC) and sol-gel process (CP with HTC) were used.

~~455 The mass-spectrometric measurements of the gas volume , within CP carrie earliet dou / showeexceedint r/8 no dU 0/ tha gt ta 2.00 1 in FM after coating deposition at temperatures lower than~~

1400 °C PCO in the beginning of the exploatation will be 20-40 ata, P thermogradienC t testa mad possiblt i e establiso t e / h/2 that at a given CP operation temperature of 1250 °C and fuel fim% burnu migratioP 15 aC o pt n wil. exceet nm l no 5 3- d Long-term heating of CP under isothermal conditions at 1800 °C gives evidenc integrite th f o e protectiof o y n coating layers. Short-term tests of CP of various temperatures indicate that damage individuan si appeaP lC r begining from 220 whic, n i 0°C s hi agreement wit date worhf th ao k /9/. Thermocycling (temperature range 350 °C - 1500 °C at a rate up to 8 K/s with the total number of cycles up to 2000) did not resul violation ti e integritth f no d initiayan l leaktightness . oCP f In different types versions of coat designs for four and five- layer CP and PyC structure composition (LTC and HTC), mentioned above, CP irradiation performance was somewhat different. The difference manifests itself ncn.attteinitial irradiation stage when fission gas release at 1000-1200^ from coatings of both versions corresponds to R/B^10 _rj , but rather when critical burnups correspondin beginnine th o gt f coago t destruction have been reached r exampleTo « t full,a y identical thicknesse protecf so - tion layers, in CP with LTC depressurization begins at higher (by 20-40% rel) burnups wit P theC hn ni ETC . Thi becauss si e radiation shrinkage and, therefore, cracking of protection layer less i sC thaLT HPCn n ni i . The results of metallographic investigations permit to make a conclusion tha performancP C t e under irradiation doet sno practically differ from that described in the literature /10/. exampln O irradiateP C f o e fim% 15 about ada t a t d 120an C 0°

456 burnu followine pth g resultpointee b n sca d out: porosit- y structur fulls ei y rearranged, pore size increased, distribution by sizes became wider and their void content increased; - metallic PP represent light bright inclusions a part of which is connected with pores; - etching in boiling nitric acid revealed grey inclusions locatingraie th n o boundariesg bode d botth an y n hi , which seem to be oxides of metallic FP not solved in the fuel matrix; - after irradiation microhardness of the fuel matrix increased

from 640±60 to 950*150 kg/mm. 2 4» Spherical fuel elements The currently adopted technological proces manufactuE P f so - ring is based on powder metallurgy methods, such as compaction with subsequent thermal treatment of blanks /11/. Refmain e I th n.2 1 result technologicaf so d experimentaan l l investigation substantiation so e VGR-5E desigP th f r no n0fo and VG-400 reactors which are under development in the USSR are presented. Because of low volume filling of CP in the two-component PE 8 vol.%(^ ) /12/ thermophysicas it , mechanicad lan l characteris- ticdeterminee sar d essentiall matrie th y xyb graphite properties. The latter, in turn, can be varied on account of initial materials (granulometrie and component composition of moulding powders). In this section of the paper emphasis is given to the influence of one of the basic factors predetermining the PE operation characteristics, namely graphite ,th e filler type. The matrix composition fillers were prepared from artificial bulk graphites base petroleun o d m cokes:d calcinatean ) PG 0 (3 d non-calcinated (MPG-6) mouldine .Th g two-componenpowdee th f ro t granulometric composition contains 0.2*0. 0.0d graphit4an n 5nu e fractions and an appropriate amount of coal-tar pitch binder.

457 Initial graphites differ essentially bot mainn hi , physical (thermal conductivity, strength etc d structuran ) e (perfection degree, porosity, crystallite size) characteristics which, to some extent, are also MG and FE characteristics. Tabl list3 e resulte sth f so investigatio maie th G nM f no E characteristicsF d an , determinin operabilite gth y e leveth f o l latter.

~~TABLE 3. THE 30PG AND MPG-6 GRAPHITE-BASED MG AND FE CHARACTERISTICS~~

~~.p . 30PG-based MPG-6-based rarame-cer matrix graphite matrix graphi~~

Density, g/ cm 1.89 1.92 Effective thermal conductivity at t » 250°C, W.m~1K-1 74 78 Specific electrical resistance 1580* 1240* at t » 20°C 1370 1150 Coefficien lineaf o t r aipansion temperature th (CLEn i ) e range 5.8* 20-1000°C, X106 degree-1 5.0 TIT

~~Elasticity modulus, X10~^ MPa 0.99* 1..P7* 0.93 1.01 Bending strengt 20°C)= t ( ha ,UP Hi 28 42 Breakin 20°C,j(« t t a gN E forcF f o e 27 39 PB wear, g 9.8 4.2~~

*The data refe o speciment r n directioi t cu s n parallel (numerator) and perpendicular (denominator) to the moulding axis.

The matrix graphites investigated are high-density carbon materials with a reasonable combination of thermophysical and physical mechanical characteristics comparable with artificial bulk graphites e samth t e A .tim e presenc f somo e e e amounth f o t second phase of non-graphitized carbon (pitch coke) results in a noticeable change in such properties as the electrical and

458 thermal conductivities, CLE, withou strengte tth h characteristics being affected. Compariso botf no G typehM s showf so thae us t high-strength non-calcinated coke- based graphite (MPG-6 type) afillesa r make possiblt si increaso et strengtE F e wead an hr resistance by about 1.5-2 times. Insignificant texture of such a filler predetermines lower anisotropy of the MG properties (e.g. CLEy ,b base)- propertiestha, G P MG d n 0 3 tha f o t t .A the same time the 30 PG based MG possessing a more perfect structure than MPG-6 has lower CLE and elasticity modulus, which is favourable for the FE radiation stability. As far as the radiation stabilit MPG-e th 6f o y materia s concerneli expectes i t i d d that it should be higher due to high strength, low anisotropy of this material and low level of radiation dimensional changes in the graphite-filler itself; currently this material is in the stage of reactor tests,

~~Conclusio. 5 n e compleTh f investigationo x s performed permitte maie th dn characteristics neede meetinr fo d e operatiogth n requiremento t s HTGR spherical fuel element selectede b o st .~~

~~References~~

1. Grebennik V.N. - Status of high-temperature gas -cooled reactor developmen USSe th Atomno-vodorodnay- R n i t a energetika i tekhnologiya (in Russion), vyp. 5., N., Energo- atomizdat, 1983 106-118. ,p . 2. Chernikov A.S.»Permyakov L.N., Mikhailichenko L.I. et al. Spherical coated particle fuel for fuel elements of HTGR.- Specialists Meeting on Gas-Cooled Reactor Fuel Development and Spent Fuel Treatment. Moscow, USSR, 1985 . 81-98,p . 3. Porthmann R. Die Chemische Grundlagen des HydrouseVerfah- renHerstellunr zu s spherischeg. r Kernbrennatoffteilchen.- Ber. Keraforschungsanlage. JU'lich, 1973, 950 s (Jul-950),35 .

459 . 4 Chernikov A.3., Ponomarev-Stepnoy N.N., Koshelev Yu.V. al .t e Spherical fuel element d somsan e result theif so r tests.- Proceedings of the conference "Gas-Cooled Reactors Today", 35-40. p , pape.87 BritisrN h Nuclear Society. London,1982. 5. Chernikov A.S., Kolesov V,S., Perrayakov L.N. et al. The constructio maid an n n characteristics of'fuel elementr sfo HTGR. Report presented to the IAEA Technical Committee on Gas Cooled Reactor, Minsk, USSR, 12-16 Oct., 1981. Eremee. 6 v V.S., Kolesov V.S., Chernikov A.3 calculatione .Th - experimental investigation HTGe th R f o fues l element construction". In /2/, p. 66-78, 7. Alekseeva I.S., Babad-Zakhryapin A.A., Glazunov M.P., TolmacheAtomnay- . N . avYu energiy Russion)n (i a , 1982, 52, vyp. 175s , .4 Khromo. 8 v Yu.P., Lyutikov R.A. Thermodynamical analysef so UOpj; - carbon interaction. - Atomnaya energiya (in Russion), 28-30. 3 , .49 . 1980t , 9. Ikawa K., Kobayashi P,, Iwamoto K. Failure of coated fuel particles during thermal excursion above 2000 °C - J. Nucl. Sei Technol.d .an 774-779. p , , 10 1978 .N , 15 , 10. Kotel'nikov R.B., Bashlykov S.N., Kaahtanov A.I., Men'shiko- va T.G. High-temperature nuclear fueRussion)n (i l , M. .- Atomizdat, . 1978s 2 43 , 11. Bedenig G. Gas-cooled high-temperature reactors - M., Atom- izdat (in Russion), 1975, 224 a. 12. Chernikov A.3., Koshelev Yu.V. Mikhailichenko L.I., Kuzne- tsov A.A. "HTGR spherical fuel elements and their experimental development". In /2/, p. 113-132. 13. Chernikov A.S., Moszherin S.I. Reahetnokov V.A.- Matrix graphite of HTGR fuel elements. Some properties. - Inter- national meeting of IAEA specialists on graphite HTGR components. Tokyo, Japan, Sept. 1986.

~~460 SELF-SUSTAINING THORIUM CYCL HIGA N EI H TEMPERATURE GRAPHITE REACTOR~~

~~K. BALAKRISHNAN Bhabha Atomic Research Centre, Bombay, India~~

~~Abstract~~

The paper contains result f studieo s r optimizinfo s g fuel utilizatio n HTGRsi n . Different U235 enrichments, core power densities and heavy metal concepts have been analyzed and results are presented and discussed.

~~. 1 Introduction~~

~~e besth Possibl tf o therma e yon l reactor syster mfo~~

~~e utilisatioth f thoriun o e pebblth kind f ms higi o be ed h temperatur cooles ga e d graphite reactor thin I .s system,~~

~~the fuel 18 in the form of coated particles, which are then~~

~~formed into balls reactoe Th . r core vesse filles I l d with~~

~~auch balls and they are cooled by helium passing through it.~~

~~The specific features which make this system attractive for~~

~~thorium ÏÏ233 e relativelcycleth e ar s y hard spectrume th ,~~

~~continuous refuelling facility, and the absence of structural~~

~~material insid e reactoeth r corel thesAl .e ten raiso t d e~~

~~e conversioth n coree ratith .n oi~~

~~The HTGR core designs thae meeton t s wit literan 1 h -~~

~~ture come e Federamostle USAth th d , y,an *& l fro e mth~~

~~Republi f Germanyo c nucleae Th . r desig n thesni s 1 e generally optimised for power generation cost, and acts to~~

~~the great detriment of fuel utilisation characteristics. The~~

~~conversion ratios quoted 5n these designs are generally of~~

~~the order of 0.85 or less. In India we have made some studies~~

~~461 in optimising on fuel utilisation. The results of these,~~

~~studie e presentear s thin i d s paper.~~

~~2. The System~~

~~The reactor system that has been considered Is very~~

~~similar to the THTR reactor. The fuel Is assumed to be In~~

~~the form of tiny particles of (ih + U233 )<> mixed oxide~~

~~surrounde sofy b d t graphit coated an e d with pyrocarbon.~~

~~Large number f suco s h coated particle e embeddear s a n I d~~

~~graphite matri d formean x d into tere sphere ar m0 1 .6 md f o s~~

~~The cor s assume 1 ee cylindrica b o t d shapen i l , wita h~~

~~diameter of 560 cms and a height of 600 cms. A conical~~

~~shaped lowe leadind en r g Int fueoa l element discharge pipe~~

~~permits spent fuel to be continuously discharged, while fresh~~

~~fue s continuouslI l fron I e top mth d yfe .~~

~~The hard spectrum In such a core results In low croee-~~

~~Beotlons and large mean free path leading to high leakage.~~

~~In the geometry described above, the leakage of neutrons Is~~

~~about 15$» In order to improve the fuel utilisation, It la~~

~~necessary to ciapture these neutrons. To do this, the core Is~~

~~surrounded by a thorium blanket which absorbs these neutrons~~

~~and raise e conversioth s n ratio.~~

~~Calculatlonae Th 3. l Formalism~~

~~This study Is intended to be a survey kind of study,~~

~~which necessarily entails calculation larga n o se numbef o r~~

~~cases* It was therefore decided to use a few group treatment~~

~~e computationsth r fo , even thoug rigoroua h s calculation~~

~~would require multigroup treatment.~~

~~462 Pour groups of neutrons -were used for the calculations~~

~~group structur s followsa s wa e : Grou s i abov kev1 1 p 82 e .~~

~~Group 2 1s 5*5 kev- - 821 kev. Group 3 la 0.625 ev to 5-5»~~

~~kev and Group 4 lies below 0.625 ev. This group structure~~

~~was assume e adequatelb o t d y accurat survea r fo ey type cal-~~

~~culation line f reasonino eTh .s followsa s gwa :~~

~~In group 1, the spectrum would correspond to the~~

~~fission spectrum n thermaI » l reactors, th«» internal spec-~~

~~trum of group 1 Is likely to stay constant from system to~~

~~system and so It Is permissible to treat it as one group.~~

~~In group 2, the primary effect is slowing down and so e Intrth a group spectrum could well depend very strongln yo~~

~~the moderator. However, in this group, all cross-sections~~

~~are quite low and fairly constant, so that even if the sp«c-~~

~~trum »thi t ver e grouno nyth s i welp l calculated erroe th , r~~

~~caused by any inaccuracies in the calculated group 2 cross-~~

~~thermay sectioan r lnfo reactor wil smalle b l .~~

~~In group 3, the over whelming effect 5s that of~~

~~reaaonances e higth hn I temperatur. e graphite reactor, with s hardeneIt d spectrum effece th , s likelfelI e tb n I to yt e thirth d group internae Th . l spectru n mgroui coulp3 e b d~~

~~different from what it is in other thermal reactors. It is~~

~~possible tha e spectruth t s morm1 ee HTG peaketh Rr fo d~~

~~towards the lower edge of the group. The third group spec-~~

~~trum will thus peak towards the lower group boundary. Since~~

~~most of the larger resonances are near the lower end of the group, this will cause the resonance absorption to increase~~

~~when compared to the normal thermal reactors.~~

~~463 decides wa t dI that some Increas e reasonancth n i e e integral of thorium 1n the third group is thus in order. To~~

~~estimate the magnitude of this increase, th^s method of cal-~~

~~culation was applied to the THTR core. Reactor physics data~~

~~was obtained from the Nuclear Reactors Directory. The date of~~

~~thp information is 1972. Although a lot of literature on THTH~~

~~has been received of late* we were unable to locate an adequa-~~

~~te amount of reactor physics data in them. The 1972 date gave~~

~~the core descriptions* well as the burnup, which was quoted as~~

~~113,000 îfcrd/T. The maximum excess reactivity was quoted as e basith f n theso sO . e11 availablmk 1 e numberss wa t i ,~~

~~decided to increase the reasonance integral in the third group~~

~~by about 15$ whicn i , h case e bot reactivite burnuth th hd an p y are fairly well -predicted .~~

~~e thermath n I l group, ther e onlar ey three nuclides~~

~~whi.ch influenc e grouth e p internal spectrum. Thee U23ar y 3,~~

~~thorium carbond an ,f theseO . , thoriuv carbod 1/ man e ar n~~

~~absorber treated an s s sucheffeca de e Th f th U23. o tn o 3~~

~~spectrum is accounted for by considering 1.t a 1/v material.~~

~~Since 1he Ü233 cross-section does not have very sharp varia-~~

~~tions, it is assumed that this approximation is adequate for~~

~~survey calculations which we have in mind.~~

~~4. The résulta of the~~

~~Calculations were made for different Ü233 enrichments~~

~~thoriumn i , different value f heavo s y metal contene th n i t~~

~~fuel balls differend an , t power densities n ordeI .o t r~~

~~evaluat e merith e f eaco t h case quantitieo tw , s were calcu-~~

~~lated each time. One was the discharge burn-up. The other~~

464 is fJaelle Inventory ratio, which is defined as the ratio between the fissile content in the discharged fuel, and that e fresth hn I fuel*

~~?ig.1 shows both discharge burn-u fioeild pab e inven-~~

~~tory ratio plotte functioa s a d f theo n . U233 contene th n i t~~

~~thorium. As is to be expected, the burn-up Increases with~~

~~enrichment, whil fissile th e e inventory ratio decreases.~~

~~-36000~~

~~HEAVY METAL CONTENT-25% POWER DENSITY-3.12t~~

~~-3opoo~~

~~-24.000~~

~~- iqooo~~

~~— 12.000~~

~~7-0 8-0 9.0 •<*233 CONTENT IN THORIUM (V.)~~

~~FIG.l HTGR CHARACTERISTICS AS FUNCTION OF ENRICHMENT~~

~~465 F1g«2 shows a plot of the sane two quantities, namely~~

~~discharge burn-up and fissile Inventory ratio, as a function~~

~~of the heavy metal content of the ball. This is expressed~~

~~In terms of the weight percent of the ball formed by the~~

~~heavy element e heavth s yA «meta l content increasese th ,~~

~~spectrum become e conversioth s hardeo s d an nr ratio Increased~~

~~At the same time, because of the decline 1n fission reactions~~

~~of the thermal neutrons, the burn-up decreases. It can be~~

~~seen that at lower heavy metal concentrations, the discharge~~

~~burn-up Increases dramatically increase Th . burn-un i e a l p~~

~~ordealmosn a f magnitudef o ro t e ligh th f f1g.2o n tI .t i ,~~

~~appears quite understandable how one could be tempted to give~~

~~6-9%' -4. 233 IN THORIUM 25-0% HEAVY META N FUEI L L BALLS~~

~~-«- POWER DENSITYtkw/lilrei~~

~~FIG.2 HTGR CHARACTERISTICS AS FUNCTION OF POWER DENSITY~~

~~466 e orospecuth p f betteo t r fuel utilisatio prefed nan o takt r e~~

~~advantag e extremelth f o e y attractive burn-up wh-~~

~~reactor syste given mca .~~

~~Fig.3 show e effecth s f poweo t r densit e disth n -yo~~

~~charger irradiation and fissile inventory ratio. The dis-~~

~~charge Irradiation comes down very sharply with Increased~~

~~parer density. This figure doe t immediatelno s y shoy wsa~~

~~effec conversion o t n ratio. That effect will however, come steo ptw a process s a . Wit decreasa h dischargn i e e burn-up,~~

~~ENRICHMENT-6-9% POWER DENSITY-3-0 kw/liter~~

~~5 3 0 3 5 2~~

~~HEAVY METAL CONTENT IN FUEL(%)~~

~~FIG.3:HTGR CHARACTERISTIC S FUNCTIOA S F MODERATIOO N N~~

~~467 an effort wil made b lo Increast e e dischargeth e burn-uy pb~~

increasing the enrichment, ani this will cause the conversion rati o comot e down*

~~• 5 Conclusions~~

~~This study seem pointo t s that conversion ration i s~~

~~e possiblar exces0 1. ef o swit e HTGRhth . Tw r threo e cases~~

~~which look promising have been selecte more, a r rigoroudfo s~~

~~analysis using multigroup codes. These cases are the~~

~~following:-~~

~~EnrichmenM H t f o % . Wt Power density (kw/Litre) % U233 bale th l n i~~

~~6.9 25.0 3.0~~

~~7.0 25.0 3.0~~

~~6.9 24.0 3.0~~

~~6.9 25.0 4.0~~

~~Prom the point of view of fuel utilisation, the HTGR~~

~~vera s yI promising system when thoriu . use mis s fertil a d e~~

~~material. The disadvantage is the rather high specific~~

~~inventory that is required if conversion ratios exceeding~~

~~1.0 have to be attained. The other disadvantage 1s the problem of initiating the self-sustaining cycle. In Indian~~

~~conditions, where enrichment facilities are not available,~~

~~the route to SSET lies via self-generated plutonium. But~~

~~plutonium is a dlsadvantaged fuel in thermal reactors. This~~

~~disadvantag s i accentuatee e HTGth Rn i dspectru m whics i h~~

~~harder than other thermal reactors, but not hard enough to~~

~~468 Teach those energies at which the <* of plutonium starts decreasing againsolutioe On . nproduco t wil e e firab l th e t~~

~~charg f U23o esann i 3 e other system lik heavea y water reactor, or a fast reactor blanket.~~

~~6. References 0 Various publication THTR-300n o s .~~

~~469 USER'S PERSPECTIVES ON GAS-COOLED REACTORS (Session G) FUTURE APPLICATION HIGE TH H F SO TEMPERATURE REACTOR~~

~~A. KLUSMANN, M. STELZER Ruhrkohle Öl und Gas GmbH, Bottrop, Federal Republi f Germanco y~~

Abstract e advanceTh d reactor systems e fas,th t e higbreedeth hd an r temperature reactor (HTR), have reached maturity for commercial e samth use et A .tim e nuclear a critica energn i s i yl situationn I . addition to reliability, safety, and competitiveness new requirements hav er futur fo com p u ee project mann i s y countries. More safety easn ,a y fuel cycle wid,a e variet f proceso y s applications, public acceptanc environmentaw lo d an e l impact e demandear e s th y b d n fulfica R publicl HT mos e f thesto Th . e criteri proves i s y a ab n the German prototype plants 15 MW AVR-Jülich and 300 MW THTR Schmehausen. The Federal Republic of Germany can contribute to the f nucleafuturo e us e r energ w safne ea y reactor system with application possibilities beyond the electric power production.

. 1 INTRODUCTION The development of the pebble bed reactor has been pursued for e Federa0 year3 th n i sl Republi f Germanyo c e maiTh . n goals have R JülicbeeAV nW M hreached 5 operate1 e :Th s since t a 196 d 7an temperature levels of up to 950 °C since 1974 and the 300 MW THTR Schmehausen is on full power. The steel vessel concept for small reactor 80-10f o s W (AVR-type0M prestressee th d )an d concrete con- W (THTR-typeM 0 60 - structio 0 ) 50 plant r e readfo n ar r commers fo y - cial use. e futurth r e Fo applicatio greaa n t variet f technicao y l options exists. Besides the normal electricity production the use of nuclear process heat of 1.000 °C allows projects from coal gasification to water fission. The development of new processes, materials, components d thei,an r adaptio e higth h o t temperaturn - re e actor is therefore subject of a program in the Federal Republic of Germany. Until the end of 1985 appr. 7 billion DM were spent on the HTR development. Thim includesu s e constructioth s n R projectAV W M 5 1 s Jülich and 300 MW THTR Schmehausen and the expenses for nuclear process heat development wor f appr o ke 4 Ministebillio Th 1, . . DM n r of Economy, Small Business, and Technology of North Rhine West- phalia, the Federal Minister of Research and Technology, and the German industry have funded the program.

. 2 ENERGY SITUATION The energy reserves of the world are not abundant and secure r ever fo e presen.Th t pictur cheaa f o ep energy surplus doet no s correspond to long-term realities. Petroleum and natural gas accoun r morfo t e tha energe n th hal f yo f consumption worldwide.

473 Large proportions have to be imported from politically unstable regions into those countries wit o majon hn sourcesl row oi e Th . price e arbitrarar s d havan y e great impacte economl th al n f o so y countries. The search for alternative energy supply systems concentrates on coal and uranium conversion to electric power and heat or gaseous and liquid products. This is a comprehensive challenge to science and technology. But these new ways are required since in the next two decades the energy consumption in the world will increase wite populationth h . Therefor e fulth el benefio ttw froe th m w materialra s coad uraniuan l m wite biggesth h t resources shoule b d reached. A long time is needed to prepare for the industrial use of new technologies for the next generation, in particular if nuclear applicatio envisages i n d beyond generatin f electricityo g e Th . optimum system that delivers different energies suc s electria h c d liquidpoweran s y interlockega ,b s d energy conversioe b o t s ha n found. The key component HTR is available, but the technologies to t developebye e t applie no r large-scal e fo d ar d e industrial use.

. 3 APPLICATION SURVEY The basic idea to use high temperature nuclear heat is to break complex natural materials suc s coa a hr heav o l y hydrocarbons into their basic components. These product e reformear s y varioub d s synthesis ways intw materialne o s wite requireth h d specific characteristics. The natural resources of coal, petroleum, and natural gas can be stretched by synthetic fuels, such as hydrogen or carbon monoxide, generated in nuclear/coal complexes. These very clean substances can be produced and used with considerably reduced environmental impact. First steps down this road have bee ne Federataketh n i nl Republic of Germany in the last 20 years. The prototype HTR's are in operation. Subsequent plants of 500 MW or 100 MW are to follow. With the next reactors the heat range of 150 - 530 °C can be served. More than half of the heat market requires this temperature level. Even for temperatures up to 750°C the materials and components are available today. For temperatures around 1.000 °C, however, more long-term experience and special processes are necessary to take full advantage. But also for this high temperature level very promising material d componentan s s qualifications have been achieved. Only mass productio f producto n s f largo justifie e eus e th s amount f nucleao s r heat. Nuclear coal gasificatio s therefori n a e major research area. The gas mixtures obtained with the heat from y coab s networklR ga d HT conversiointe fe e d th oe th an sb n ca n used as fuel gas for home-heating or for industrial processes. Also synthesis gas for the production of synthetic materials, ferti- lizers d liquian , de producedb fuelr botn fo ca sht purposeBu . s industrial plants are not yet under consideration. Further proposals concern the steel industry, such as nuclear smelting of iron ore. A switch from blast furnaces and coke to shaft furnaces is envisaged in these proposals. - Also the cold

474 transpor y becom w areama f hea o tne e converR a tTh e.HT froe -th m sion of methane into hydrogen and carbon monoxide and then in the place of consumption the conversion of the fission gases back to methane are proposed. - Also the noble form of energy, hydrogen, can be produced by coal or heavy hydrocarbon gasification or later y direcb t water splitting. Thus the pebble bed reactor application to the heat market make industriath e l countries more independent from energy imports. The HTR can contribute almost all energy qualities, but there is a long technical way until these routes will be competitively established. Thia heliu f so applieme gasturbinus e sth alsr n fo oi e direct cycle.

4. PROJECTS THTW M R0 Th Schmehause30 e n represente th e basi r th s fo s commercialization of the HTR. It is the modern German reference plant for all subsequent projects in the power range between 100 e extensivTh . MW e0 an 60 desigd n reserve e reduceb n ca sd with growing operating experience on the way to competitiveness. By a twin design with one secondary part a power range up to 1.200 MW can be covered. The next step of commercialization is a 500 - 600 MW steam cycle plant for electric power generation with the possibility of process steam and district heat extractions. Industrial reactors designed for 100 MW power output and as multiple stations for 200 and more MW make use of the proven principle of the 15 MW AVR Julien steel vessel design as well as the THTR-technology e Federath n I .l Republi f Germano c o reactotw y r supplier e competinar s n thii g s area wit e pebblth h d designbe e n I . the United States a bloc fuel reactor of slightly larger size is under development. The lay-out difference concerns the integrated and non-integrated arrangement of the steam generator. These proposal e suitable supplar sth f energr o y fo e o industriat y l com- o plexesmalt d l an s gridr countrie fo s wel a ss a l s with less developed industrial structures. e R+D-prograTh e conventionath r fo m l lay-oue HTR-plantth f to s is retrogressive concernin e publith g c funds e planninTh .w ne f o g HTR power e financeb statione privato th t y s b dha s e organizations in the Federal Republic of Germany. However, the development work f higo e h us temperatur e th r fo e a heasubsidize s i t d programme known under the name "Prototypanlage Nukleare Prozeßwärme" (PNP- project) wite participatioth h f severao n l German organizations. To meet the long-term objectives of the high-temperature proces e absolutsth hea e us t e prerequisit e realizatioth s i en a f o n intermediate HTR for cogeneration of electricity and steam. The poin s beeha tn reache y studieb d s thae presenth t t designe ar s feasible, licensabl d almosan e t competitive r furtheFo . r technical progress a follow project is needed to demonstrate serial production, convoy construction, and other ways to reduce costs by a simple t fullbu , y safe plant lay-out.

475 . 5 PROCESSES UNDER DEVELOPMENT The potential for energy systems with high temperatures is in several industrial large-scale processes, such as hard coal and lignite gasification, iro r hydrogeo n n production. 5.1. Steam Gasification of Hard Coal To take full advantage of the HTR heat a new gasification e developedb proces o t d r thiha s Fo . s purpose from 197 o 1986t 4a semi-technical fluidized bed gas generator was in operation at the Bergbau-Forschung GmbH in Essen. The gas generator with a coal e tesTh t . °C 0 80 o t 0 70 d an r operateh ba t/ capacit0 4 2 t 0, a d f o y runs had the objectives to demonstrate that the heat demand for hard coal gasification can be supplied from outside. Improvement of e conditionth e fluidizeth n i s d bed e hea,th t transfen a y b r immersion heater, the tests of refractory materials, dust separators, heat recovery systems, and high temperature fittings were the test targets. Within a total operation time of 26,000 h, all objectives were achieved. About 1,700 t of anthracite and semi-anthracite coal were gasified. Finally, it was possible to gasify also baking high volatile bituminous coal (70 ) t afteinjectio0t je ra n syste s installedmwa . After the test runs an engineering assessment will be available by the end of 1986. The evaluations indicate so far the technical feasibilit e steath mf o ygasificatio n - procesin r fo s dustrial application, wherea e competitivenesth s t givenye t .no s i s e revieweIth n d concep a theliu m cycle wit a capacith f o y 340 MW(th) at 900 °C is adjoined to a gas generator. A horizontal typs generatoga e r wite immersioth h n heater concept would hava e length of 33 m and a diameter of 7.2 m at a capacity of 28 t/h and a gasification rate of 93 %. Boosting the gasification reactions by adding a catalyst could increase the capacity to 69 t/h. - Also a preliminary design of an industrial hydrogen production plant is performed. The concept includes three He-cycles and gas generators connected to a 1,000 MW(th) HTR. To increase hydrogen production the installation of a helium heated steamreformer is planned. A demonstration plant would have a coal throughput of 0.6 mill, t/a to produce 2.0 billion m3 hydrogen. There is further development potential if high temperature materials can be proven for catalytic gasificatio w gasificatione r o n n concept e consideredar s .

5.2. Hydrogasification of Lignite A hydrogasification process wit a verticah l fluidized be d reacto s undei r r developmen r lignitefo t . Hydrogasification a s i n exothermal reaction. However, HTR heat is necessary to reform the methane produce e gasifierth n i d e reformeTh . s heatei r y heliub d m gas of 950°C. The produced synthesis gas is processed to hydrogen which is fed back to the hydrogasifier. e successfuTh l test a semi-technicarun f o s l plant a t Rheinische Braunkohlenwerke AG in Köln-Wesseling permitted the scale-up to a 10 t/h-pilot plant which is in operation since May 1983. The experimental test runs were finished at the end of September 1986 n tha.I t time planth e t operate r 14,72fo d 3h including 7,77n coalo h 7 . More than 38,10 f lignito 0t e were

476 gasifie 9 millio 1 f methane o o t d 3 m ne methanTh . e concentratiof o n gasifiea t ranges a ga % w thr6 d 3 ra eefficienc betwee d an 2 2 n f o y . Unti% 0 l6 o Aprit 0 l5 198 7a detaile d documentation wil e conb l - cluded. The gasification process for lignite is due to the pilot plant operation far ahead of the gasification of hard coal. However, no actual pla o scale-ut n e procesth p presentls i s y madr economifo e c reasons. 5.3. Materials Program Froe beginninth m e developmenth g e materialth f o td compoan s - a crucianent s swa l issue w alloyNe . s suitabl o procest e s coat a l high temperatures in an eroding and corroding atmosphere were necessary. Until today material scientists are still testing new resistant qualities especially for the catalytic process conditions e heaw fielth ne ts i A exchange.d r technologo t 0 80 t a y e prograTh r . hig1.00fo °C m h0 temperature material w compone d -an s nents such as the He/He-heat exchanger, the high temperature helium loop, the reformer, the hot gas tubes, and the fittings requires up years5 t1 o . Component modules have been e testehigth hn i dtemperatur e helium loop facility at Interatom GmbH in Bensberg. Since August 1982 test condition t temperaturea s o 1,000°t p pressured u s Can p u s to 46 bar in steady or dynamic mode have been available. Until September 198 e tes6th t facility operate r 11,00fo d 0h includin g 3,300 h at temperatures higher than 900°C. Test items are the hot- gas pipes and fittings and other crucial new designs. The efforts are concentrated also on carbon compounds and ceramic high temperature insulation materials. Many important results have been achieved. The He/He-heat exchanger and steamreformer materials are qualified for 60,000 to 100,000 h. For high operation temperatures "Inconel 617" and for lower temperatures "Hastello " mee X e yrequirements th t r fo t bu , catalytic gasification conditions the material problems are not yet solved. The feasibility to produce largesized forgings has been demonstrate t collecto ho y fabricatio He/Hee b dt th 1 1 f - o ra f o n heat exchanger. - For hot gas lines at temperatures higher than 600° w carboCne n compound tubee s ar wit m m diametea h 0 90 f o r available. - The future development activities are directed towards industrial applicatio e materialth f o nd componentan s s testedt I . is expected tha e materiath t l development period will last until e latth e 1980' r eveso n longe ceramif i r c material e alsar s o considered. 5.4. Iron Production The HTR is intended to be used to supply heat and steam to gasify hard coal up to 50 % to synthesis gas. The residual coke is n iroa d bac nfe o t k bat r reductiofo h f oreo n . Wite Klb'ckneth h r steel-gas-process coke, oxygen, and iron ore are processed in an iron bath. The synthesis gas generated can be converted to methanol. The products of the integrated plant are electrical power, methanol g iron.pi d Untian , o actua w n ther no e l lar e plans for an industrial plant. Before industrial use the technical feasibility of all production steps has to be demonstrated.

477 5.5. Hydrogen Production Hydrogen production by using electrical power and heat is a further HTR-application, which could become very importane th r fo t future energy market. Conventional electrolysis processes utilize electrie th f o nearl c% power0 8 y e efficienc.Th f thio y s process with light water reactors is approx. 26 %, whereas the HTR may . Efficienciereac% 1 3 h f moro s - e, howeverex % tha e 5 b 4 n n ,ca pecte y melb dr higo t h temperature electrolysis, becaus a pare f o t e e energintroduceb th n ca y s heata d . Electrolytic-chemothermal hybrid processes with sulphuric acid separatio t temperatura n e level f 400-50 o 800-900°sd an C 0° e furthe Car r possibilities.

6. MARKETS 6.1. Electricity/Steam Cogeneration electricits i R HT e th ye primar f Th productioo e us y d steaan n m supply with units up to 600 MW. The competitiveness in this market against light water reactors has to be proven, but is a realistic assumption. The electricity market will be open for the HTR when economic proposal e reacto th e mad y ar sb e r industry. In areas with medium-sized grids small e user b HTR' fo dn ca s electric power generation, steam production, and community district heating. Alsr thosfo o e countries wit o nuclean h r e prograth t ye m y becomma smal R a vereHT l y safe option n countrieI . s withou- in t dustrial infrastructure small reactor n alsca s o serve certain island region r deliveo s r e high-pressurth o t C ° 0 53 e o steat p u m industry. Energy supply for tertiary oil extraction, processing of heavy oil and tar-sand or for desalination, and coal gasification e fieldth r applicatione fo sar . Recently the IAEA identified in detail the potential markets r smalfo l nuclea e advantageTh r. poweMW 0 sr 30 quote planto t p du s are: Lower absolute capital cost causes a smaller financial burden and reduce e economith s c risk e earlTh . y introductio f smalo n l nuclear power plants reduces the environmental pollution. Finally small plant e suitablw capacitar slo r fo ey grid d applicationan s s r selectefo d industrial processes e IAE.Th A expects that some 5 additiona2 l countries have grids wit capacita h f 2.00o y 0- perioe th 6.00f n 199o i d W 0M o 200 t 0 wit markea h t a read r fo y number of small nuclear plants. In countries with large interconnected grids, high technical infrastructure and energy consumption the market for the HTR comprises the whole electric power and heat supply. Nuclear power plants in those countries have sizes between 500-1.300 MW(e) for economic reasons. Smaller reactors would not be competitive against large nuclear plants. In those areas with a well developed technical infrastructure and high industrial production a wide field of new applications for nuclear steam is open. Low temperature heat is require a numbe n i d f industriao r l branches e chemical, th suc s a h , paper, cellulose, sugar, textile industry or for district heating. But in almost all cases the required amount of heat from a nuclear reactor is limited in each site. Steam supply therefore has to be coupled with electricity productio r smalo n l HTR-plant f somo s e e builtb o t . e ar W M 0 10

478 6.2. Situatio e Federath n i nl Republi f Germano c y Ia timn f moderato e e growth rate f energo s y consumptiow ne n sales can only be made at the expense of competing energies. The prospects for the introduction of HTR in the Federal Republic of Germane regardeb n exampln a ca ya typicas a df o e l industrial country. In the Federal Republic of Germany nearly 95 GW power station capacit installeds i y , which produce 0 TWh40 s A forecas. f 198to 4 predicted for the year 2000 nearly 110-125 GW capacity to produce 480-560 TWh. These values may be lower considering the power plant structure. But from 1995 several GW per year additional capacity is required. Some HTR with 500 MW could be accomodated from the market poin f viewo t . Another example is the German market for synthesis gas. The chemical industry needs about 10 billion m3/a. The conversion of 0,6 Mio tee (maf) hard coal delivers up to 2 billion m3 of hydrogen and carbon monoxide using nuclear heat froe HTR e totath m Th . l annual deman e producef b f synthesio o dn e ca te s d o ga sfro Mi 3 m hard coal with heat supplied by two large HTR's. This market could expand considerably in case the methanol sales raise for fuel supply. This market section wil e determineb l y competitivenesb d s with minera l productsoi l . Three percent methano e usuath n li l fuel mean n additionaa s l sal f methanoo e o t/alMi fro.7 A complet 0, m e substitution of the gasoline market by methanol would require the additional - amounun s f methano i o tr yea d pe an rt lo Mi fro 5 3 m realistic. e existinTh g heat markee Federath n i t l Republi f Germano c s i y divided into the low temperature range and the process steam supply for industry and small consumers. Forecasts assume 180 Mio tee in the year 2000. Oil saving potentials and increasing demand represen tteeo apprMi .5 Sufficien4 . t potential exists thereforr fo e the introductio f nucleao n r heat. Toda e proportioth y f districo n t heatine te o apprs i gMi 6 . rising to 12 Mio tee in 2000. A detailed study about the application of district heat concludes that in nearly 190 cities existing heat demand could be served by district heating. The main demand is in urban centers to be connected with power stations. Assuming a grid enlargement nuclear power stations could substitute appr. 6 Mio tee per year competitively if a base load of 7.300 MJ/s and 6.500 h of operation were presumed. At the same time the nuclear plants coul de additiona producth f o % e l0 betwee1 n a 5 n electricity demand unti yeae th lr 2000 e introductio.Th f nucleao n r district heat depends on the availability of large grids. At the moment the introduction of nuclear power plants for this purpose is a matte f singlo r e cases A maximu. f fouo m r power plantr fo s district heating can be considered until the year 2000. The demand of industrial process steam energy will reach 2000n i e . te Thi o sMi include5 6 modesa sr pe % t 4 growt0. f o h year. The main part of the process energy is distributed amongst the following industries with two crucial temperature peaks: Up to : chemical°C , paper/cellulos0 % 40 5 2 s sugad , textil% an 0 r% 1 e 4 e 2 %; at a level between 900 and 1.700 °C: steel industry 56 %, cement/brick 29 %, glass/ceramics 8 %, chemicals 6 % and aluminium 1

479 Wits speciait h l advantage o product s e electricit d procesan y s steam simultaneousl n serv ca e chemica th eR HT e l th y industry first. The energy intensive processes for producing chemical feed stocks e large-scalar e operations wit a highd constan an h t operation rate over the whole year. The industrial application of high temperature e industrsteath n i m s als w i singly fe mattea o a ef o rcases . Nevertheless, because of the existing infrastructure this market segment is very important as a first application. Although long-term development programme e underwaar s s i t i y still too early for a prediction of the future market for HTR high temperature heat for coal gasification, steel production or water splitting.

7. COST AND SCHEDULE Whereas the HTR use for electricity and steam production is a commercial matter, the introduction of technologies around the HTR is determine e requirementth y b de futur th f eo s market r gaseoufo s s and liquid products. Also the availability of special processes, and the political and industrial determination for new projects pla majoa y r role. Furthermor e introductioth e f advanceo n d systems and products depends on their development costs and the prices for other energies. As a first guideline for the economy of the new processes coal gasification costs can be used. Synthesis gas of modern autothermal coal-dust gasification plants costs around 0.22-0.24 DM/m3 with feed coal prices of about 200 DM/tce. Economically less favourable is the production of substitute natural gas estimated to cost

0.75 DM/m, more than twic mucs a es importe a h d natural gas. 3 e nucleath r rFo coal gasification production cost estimates had been calculated in the early 1980's. At that time appr. 70 % higher costs were assumed thae naturath n s imporga l t pricf o e 0.30 DM/m3. Recent calculations show that since then no significant improvements towards better economy have been achieved. Moreover, more development time is required than originally anticipated, which means, that the present low oil and gas prices might have increased again by the time the technology will become available. e e advanceoptionth Th r fo sd HTR-concept e maintaineb n ca s d only if the 300 MW THTR Schmehausen operates with full capacity for several years and if the follow-on projects HTR 500, HTR 100 (BBC), or the modular HTR (KWU) are successful. Many long-term experimental steps with public funding are necessary. The project schedule requires time unti e nexth lt century, becaus e chanceth e o t s connect a new application route to a HTR reactor have to be regarded as open or very long-term. Also for example for the hard coal gasification with nuclear process heat very high development costs will arise. First estimates for a nuclear test of the new gasification process , althougarrivDM o t approxa e Mi a successfuh 0 50 . l programme of 1.4 billion DM has already been performed in the past years. Therefore for the future programme the PNP-project might encounter some delay.

480 8. SUMMARY The HTR for electric power and steam (530 °C) production is feasible and almost competitive for plants up to 500 - 600 MW. Smaller reactors are under development. Concerning the advanced application e transfeth s a heliu t f heaa o r mR t outleHT fro n a m t r processefo temperatur C ° 0 s 95 suc f s coao ea h l gasificatios i n feasible. The work on the nuclear island for future plants is underway. The new industrial concepts take into consideration the option for the future use of the high temperature range. The tests of the heat transfer components and the selected materials for temperatures of 950 - 1.000 °C are successful, but continuatio r somfo ne more year requireds i s . Hydrogasificatiof o n lignite is more advanced when compared with the steam gasification proces f haro s d coal e futurTh . r botfo e h concept opens i s . Follow- on plants are not yet decided. By the end of 1986 an engineering assessment will be completed providing the basis to discuss the next project stage. All other technologies are considerations for the far future. The decisions ahead will be difficult to make. At a time of scarce funds for development, very low oil prices, and a general distrust in nuclear energy application the long-range goals of the HTR-application have to be reassessed. Since the present energy situation does not justify the necessary expenditure for private companies only further public suppor abls i to maintai t e e th n incentives for further advancing the described projects. A more general view on the world energy situation is advisable in this situation. It is assumed that increasing world population and demand in developing countries will lead to more energy consumption. Representatives of the oil industry regard the present pricl oi ew lo leve s temporara l d assuman y e tha e worl th l prot oi d - ductio d demanan n dn balanc i wil e b l e again latese 1990'th n t i tsa a higher price level. Furthermor w politicane e l problems cannoe b t excluded. Another reason to develop new technologies is the ratio f fossio l energy reserve consumptiond an s e fossith f .lo Abou% 5 7 t world energy reserves are coal and 25 % oil and natural gas. The consumption data are almost the reverse: oil and gas 60 %, coal . Therefor% 0 3 e clean technologie r coafo sl upgradin e needear g d and the use of nuclear heat could well be the solution. e contributioTh f nucleao n r worldwide energth o t y e energy % onlsuppl 5 ys i yalthoug 0 nuclea35 h r power station open i -e ar s ration. There is a certain hesitation to proceed with new plants afte e accidentth re Uniteth n i dse cos Russin Stateth i t d d an aan s increase for nuclear power in the last years. The HTR has safety advantages, application possibilitie w processene r fo d sstand an s s in competition wite establisheth h d light water reactors n thiI . s situatio w generane a n l discussio o proceet w ho nd long-tery ma m arise but requires time. In any case a next HTR-project is needed soo o avoit n a scatterind e capablth f o ge engineerin d researcan g h teams built-up with billions. A planning order for the HTR 500 is the most advanced next step and also the studies for smaller HTR plants could contribute.

481 The chance to choose an alternative nuclear system is given, since the 300 MW THTR Schmehausen has reached full power. The use R besideHT e th s f poweo r generatio w materialra r fo n s production might determin markes it e t introduction e frequentlth r Fo . y announced high application potential the HTR community needs an international cooperatio f variouo n s branche energe th f yo s industry. Coal and nuclear energy are the future resources, and their combined use is a common long-term task. The continuity of the development e subjecshoulb t o no evedt r changing price cycles energe th f yo market rathet consequena ,bu f o r t energy policy.

~~. 9 REFERENCES~~

(1) BARNERT, H., SCHULTEN, R.: Der Hochtemperaturreaktor. Jahresbericht 1984/85 der KFA JUlich GmbH. (2) VAN HEEK, K.H.; JÜNTGEN, H. und PETERS, W.: Wasserdampfver- gasung von Kohle mit Hilfe von Prozeßwärme aus Hochtemperatur- Kernreaktoren. Atomenergie/Kerntechnik ^0 1982, 225. ) KLUSMANN(3 , SPECKSe HigA. ,Th h : Temperatur,R. e Reactod an r Coal Conversion (VIII. International Conferenc e HTGRth n )o e San Diego, 15.-16. Sept. 1986. (4) KNIZIA, K.: Entwicklungsschritte in neue Energiekonzepte. Kohleumwandlungstechniken und Hochtemperaturreaktoren für den Energiemarkt der Zukunft. VGB-Sondertagung, September 1985. (5) KNIZIA, K., SCHWARZ, D,.: Der HTR 500 als nächster Hochtempera- turreaktor KraftwerkstechniB VG . ^ (19856 k ) 195. (6) MESSERSCHMIDT, H.: Die Bedeutung nuklearer Verfahren für die Kohleveredlung. Atomwirtschaft August/September 1983, 467. ) NICKEL(7 , SCHUBERTH. , , PENKALLAF. , , H.-J.: Stan r Qualide d - fikatio r Werkstoff de nn HTR de ,r VGB-Sondertagungfü e , September 1985. (8) TEGGERS, H.: Vergasung und Verflüssigung von Braunkohle. Braunkohle 3_7_, (9) 1985, 390-393. (9) SPECKS, R.: Die Chancen der Kohlenvergasung, Glückauf 114 3^ (1978), 137-142. Entwicklungsstand der neuen Technologien der Kohlenvergasun d Kohlenverflüssigungun g , Glückau3 2^ 9 11 f (1983), 1147-1159. (10) SCHULTEN : GelöstR. , nocd u lösendun ez h e Problemr de i be e Entwicklun n Hochtemperaturreaktorevo g n Verfahrevo d r un n zu n Nutzung der HTR-Wärme, VGB-Sondertagung, September 1985. (11) SPECKS, R.: Kohle und Kernenergie, Wintertagung Deutsches Atomforum e.V., Januar 1986. (12) HKG GmbH: The Other Way of Using Nuclear Energy, Mai 1986. (13) Arbeitsgemeinschaft Hochtemperaturreaktor, Interne Dokumenta- tionen R 500-Vorprojektuntersuchungen HT R Modu d HT : un l , 1984. (14) VDEW e.V. e volkswirtschaftlich:Di e Bedeutun elektrischer de g n Energie auf dem Wärmemarkt, Frankfurt 1980. (15) Arbeitsgemeinschaft Fernwärme e.V. (AGFW), Hauptbericht Fern- wärme 1982.

~~482 FINANCING MODELS FOR HTR PLANTS Co-financing, counter trade, joint ventures~~

~~J. BOGEN, D. STÖLZL Brown, Boveri und Cie AG, Mannheim, Federal Republic of Germany~~

~~Abstract~~

Structur d voluman e f investmeno e R tHT cosr fo t nuclear power plant e differenar s n comparisoi t o t n other types of nuclear power plants. Even if the share of local participation is in comparable order of magnitude to other nuclear power plants, the required technical infrastructur R plantHT r s i sfo e more suitable for existing and still practised technologie countrien i s sn developmen i whic e ar h t processes. These HTR specific features offer special possibilities in HTR project financing.- Various models are discussed in respect of the the special HTR situation. Even if it is not possible to point out in a general manner the best solution - due to national, local and time dependant situations - this paper discusses the HTR specific impacts to buyer's credit financing, supplier's credit financing, barter trade r joino s t ventures and combined financing.

. 1 GENERAL In general financing of nuclear power plants is characterize y higb d h capital requirements in form of investment cost and low cost for operation, maintenance and the fuel cycle |l|. Nations which desire to use nuclear energy usually have to ratify some international agreements start energy demand analyse accordancn i s e to populatio d economian n c growth rates start economic analyses on different energy generation alternatives (water, oil, gas, coal, nuclear) start technical know-how improvements start actions for improvement of the general infrastructure basis evaluat e environmentath e l situation.

483 Onl f suci yn (independenta h ) assessmenr to feasibility stud ye resul cometh o t s thaa t nuclear power plans i t viable desirable for the foreign country, it makes sense to discuss financing models. n thiI - s s papei t i r not the task to create new financing models for HTR-plant o shot wt bu stha t HTR-plant e morar s e convenien r financinfo t g than nuclear power plants (NPP) in the range of 1300 MWe. . 2 Financing nuclear power Nuclear power project e characterizear s y b d years5 o t '0 1 perio a f plannino d g prioo t r the star f constructiono t e constructioTh . n P takeNP e sperio th abou f o dyears6 t . After handove e operatioth r n e planperioth s i f to d y witsa hn ca 0 years. e orde3 e ith non f o ro -N the confidence requiree assessmenth r fo d f o t financing risks what will happen to the politi- l stabilitca r economio y c strengt a countr f o h y (under development) over suc a timh e periof o d years0 5 4- 0 . Therefore, one of the first prerequisits for financing a nuclear power project is to shorten the construction time of the plant and to minimize the pay-back period for credits and financing cos y saleb t e energf energo sth o t yy consumers such as industry, trade, private consumer eventualland s exporby y electriof t c energy. This requires a proven technology of the NPP technology decider fo d an experienced supplier's management during construction a well trained operation personnel guaranteed earnings from sales of energy economic and financial conditions to convert earings from energy sales from national into international currency. e HTR-linTh e with spherical fuel elementf o s graphite and the inert gas helium as primary coolant represents a proven technology with high safety standards and good-minded operation characteristics. This had been demonstrated by the AVR-15 MWe experimental power plant, n operatiowhici s i h n in Jù'lich, (F.R. Germany) since nearly 20 years. An experienced supplier's management

484 coul e createb d d during constructiod an n commissioning of THTR-300 MWe prototyp power plant in Hamm-Uentrop, (F.R. Germany). By implementatio f thio n s project alse th o required infrastructure for licensing and sub-supplier's supplies and services was built-up. The THTR-300 MWe plant is the first nuclear power plant in a technical scale worldwid e pebble-ben basith o e f o s d principle. On Sept , 198623 .THTR-30e th , 0 generate% 0 10 d thermal capacity. That means tha prototypa t e plana pebble-be f o t d reacto a technica n o r l scale is available as reference plant for this reactor line.

2.1 BUYERS CREDIT BASIS FOR EXPORTS FINANCING In this e plan case buyeth th etf o rask a s e exporter'th ban n i k s countr r credifo y o t t e planth ty suppliepa r accordin e supplth o yt g contrac d projecan t t progres a cas n ho s basis (fig. 1). Raw materials Official authorities: Ministr f finance/economyo y Central bank Commodities

~~+ 20+18+16+14+12+10+8 +6 +4 +2 0 -2 -4 -6 -8-10 Years~~

~~Money (internationally convertable) _-_- Local money •C=MC Raw materials, commodities - - — Function (unfillee b o st d~~

~~Fig. 1: Export Financing on Buyer Credit Basis~~

485 For follow-on projects BBC/HRB concentrated it's interes o plantt t f mediuo s d smalan m l capacities, such as HTR-500, HTR-100 and s wel a s GHR-1 a l0 HTR- 10 0x 2 (th) o t |2| e .Du the fact that thes e loweth er o n rplanti e ar s medium capacity scale they meee th t requirement f smalo s d mediuan l m sized networks and the absolute amount of the investment cost n Deutschmar(i k (DM) e ordeth )f o n ri onl s i y 17 x 106 [DM] for GHR-10 (th) 630 x 106 [DM] for HTR-100 900 x 106 [DM] for HTR-2 x 100 6 [DM10 r HTR-50x 169fo ]0 0 This is demonstrated in fig. 2. The figures are only for demonstration in the order of magnitud r investmentfo e e mentioneth f o sR HT d plants casn i ,f sales/contract eo s they havo t e be adoptee actuath o t ld e datd scopth an e f o e plant supplier.

~~x10~~

~~GH(th0 1 R)-~~

~~HTR - 100 MWe~~

~~e MW 0 10 x HT 2 R-~~

~~100 200 300 400 500 600 700 800 900 1000 1100 1200 Plant capacity [MWe]~~

~~Fig. 2: HTR-Power Plants Investment Cost~~

486 HTR-500 The absolute amount necessary as credit for financin n HTR-50s onla i f y| o g abou|3 0% 0 5 t of the amount for a large 1300 MWe plant. This reduction of the absolute investments is e econome generate th e costh th f f o t noo y y b t d nuclear power. Due to specific characteristics of HTR plants such as high total efficiency and high burn-up of the nuclear fuel (110.000 MWd/t HM compared to 35.00 0r PWR's MWd/fo e electricitM H t)th y generating cost of an HTR-500 are comparable to those of a large 1300 MWe plant. This makes finacin n HTR-50a f o g 0 easier, assuming that normally ) DM 6 10 x 0 34 = ( % 0 2 are financed by local sources/credits and the remaining ) DM 6 135= ( 10 8 % 00x can be granted by export-credit agencies to a value of 75 % (= 1268 x 106 DM), according to present condition e OECD-Consensuth f o s s a s agreed upo n Augusi n t 1984. The remaining 5 % (= 85 x 10^ DM) then must be financed by commercial loans. As any commercial bank is likely to limit its participatio e projecon r o aboufo t tn t 50 x 106 DM or less, there are still two commercial banks necessary for financing the remaining share, in the case where also the e b o t s ha ) DM loca 6 10 % (34l x 0 0shar2 f o e financed about 10 commercial banks have to be linked together for such a project. This is only hale numbeth f f banko r s necessarr fo y financing of a large 1300 MWe plant. This helps to reduce financing cost for HTR projects. The main difference between commercial loans and export credits is the grace period, that is the period before repayment of the debt starts. For commercial loans this grace eth perion i s i d order of 4 - 5 years whereas for export-credits this period is in the order of 7 - 10 years. This means that for commercially financed shares repaymen e debth t f o tstart s before th e plant delivers earnings by energy generation. The investment cost of HTR-plants for heat and power generation can be roughly characterized, as follows % 0 4 e nucleabelonth o t gr heat supply system (NHSS) and turbogenerator

~~487 and 60 % belong to sub-suppliers for the e plantth res f o .t This is demonstrated in fig. 3 for an HTR-500 plant.~~

~~100% = 1690 Million DM~~

~~: 3 HTR-50Fig. 0 Share r Supplierfo s s~~

This give n opportunita s o participatt y e local industries during the construction of the plant. The share which local industry can take depende existinth n o s g industrial infrastructure e factTh . , that an HTR-50 a reacto s ha 0 r pressure vessel of pre-stressed concrete which has to be constructed, isolate e d testeth an d t a d site e componentth s suc s steaa h m generators, valves, piping (including supports/hangers) isolations are of smaller dimensions than for large 1300 MWe plants the turbogenerator of a conventional type as required for fossil-fired power plants with comparable capacity indicates that the share for local industries e firsisth evetr fo HTR-50n 0 plant abou% 0 4 t or more. This is a greater number than normally is achieve r largfo d e 130 e plantsMW 0 .

~~488 HTR-100 For the HTR-100 and HTR-2 x 100 the advantage of low total investment cost (see fig n ) leadincreas.a 2 e o t th s f o e specific investment cost (see fig. 4).~~

~~«6115 HTR-100 HTR-500~~

~~DM kW~~

~~6000- HTR-502X 0~~

~~5POO-~~

~~4000-~~

~~3000-~~

~~2000-~~

~~1000-~~

~~0-J—A—r 100 200 300 400 500 600 700 800 900 1000 1100 MW~~

~~Fig. 4: Specific investment Cost for HTR-Power Plants~~

This influences economic figures, so that the competitivness for power and heat generated n HTR-10ba y 0 (HT x 100 R2 ) only existn i s comparison to fossil fired plants and no longer for large nuclear power units. But HTR-100 plants are of high interest in cases to avoid dependencie n importeso d fossil fuels (oil, gas r powefo ) r generation investments for an additional infrastructure r transportatiofo . g . e , n of coal investments for stack-gas cleaning systems (desulphurisation, denitrification).

489 e sharTh e which local industr n takca y e during implementatio n HTR-10a f o n 0 plans i t comparabl e HTR-50th f d o 0an o tha t e e on t depends strongly on the technical infrastructure of the country in which the HTR-100 plant wil e erecteb l d (see fig. .5)

~~Millio0 10063 M nD = % : 5 HTR-10Fig. 0 Share r Supplierfo s s~~

e HTR-10Th 0 normall a pressur s ha y e vessel of steel which is pre-manufactured, tested and pre-installed with the reactor internals at the workshops of the plant supplier. This helps to reduc e constructioth e n perio years4 o t d , commissioning included. a verThi s i ys important factor for the financing point of vied contributean w w financinlo o t s g costr Fo . a n% shar HTR-10 0 f loca2 o e e lth 0 credits a - s e assumeTh . DM d 6 befor10 x amount- e6 12 o t s 75 % share financed by export-credit agencies amounts to 472 x 10^ DM and the remaining 5 % share which e needfinanceb o a t s n o d commercial credit basis amounts to 6 DM.10 e locax - th Eve2 3 f li nshar e needs financin y commerciab g l loans, ther e onlar ey three banks necessary to be linked together, each one with a participation of 50 x 10^ DM.

GHR-10 (th) As far as the gas cooled heating reactor {GHR-10 (th)) is concerned, the total investments are only in the order of e nucleaTh . DM r 6 par f 10 thio t x s 7 plan1 t is completely premanufacture e supplierth y b d . This means that the local industry only can participat e civith n e li eth work d an s installatio e districth f o n t heating network which the GHR-10 (th) serves. The construction e planperioon s reducei r t fo d o onlo t d tw y years in total, commissioning included.

490 Financing of about 17 x 106 DM of investment cost is not a general problem. If the local % thishar 0 s2 assumes i ee b o t d shar% 5 7 e e Th . DM correspond 6 10 x 3 o t s finance y export-credib d t correspond o about s t e loca th . Eve DM lf i n 6 shar10 1ex 3 requires for credit a commercia n o s l basis, this corresponds together wite remaininth ho t % 5 g about 4 x Id6 DM. Due to the short construction period these amounte easilb n yca s financed by one single bank. e onlTh y proble s tha i mo license n t d siter sfo such plants are available and often the infrastructure for a (expensive) district heatin e erectegb networo t ds ha beforek . Above we discussed the advantages of the various HTR plants on basis of a buyer's credit and cas e planth th mone r -fo y supplied an r sub-suppliers. Today this will be something like an ideal situation, more convenient to demonstrate the advantages of HTR-plants from the financing poin f vieo t w thar realistifo n c project financing. t onle amounno th y s Ii f moneto t y which hinders the financing of nuclear power plants but alse credith o t worthines f countrieo s s a s it is seen by international lender organisations. For the assessment of the credit worthiness e debth t situatio e buyer'th f o ns country the compariso e leveth f deb o d lf o nan t debt service against the cross national product the suppl f foreigo y n exchange through expor f goodo t s e maiarth en indicators. Ther e countriear e s which have reschedule e reimbursementh d f o t thei e rinterese generallth debtar d e an sdu t y payed step by step. In this situation o buyer'thern s i e s credit granted for nuclear power plants. Then other financing models are possible which have their justification in the interest of the plant supplier and the economic policy of the plant supplier's governmen o promott t expore th e f o t power plants |4|. 2.2 SUPPLIERS CREDIT FOR EXPORT FINANCING in this case the plant supplier asks for a credit on basis of the signed contract (see fig. 6).

~~491 Raw materials Official authorities: Ministry of finance/economy Central bank Commodities~~

~~20+18-1-16+14+12+10+8 +6 +4 +2 0 -2 -4 -6 -8-10 Years~~

~~Money (internationally convertable) -__- Local money -cra materialsw cRa , commodities - — — Functions to be fulfilled Fig : 6 Expor. t Financin n Supplieo g r Credit Basis~~

For minimizatio e credith f to n amount necessary there is a certain interest to increase the participatio buyee th rf o n (high local participation o split r )o t financing betweea n large numbe f sub-supplierso r . Each party then is responsible for financing of his scope (multi-national financing approach). In this case the foreign buyer remains responsibl e planth r t fo eoperatio d payan ns bace credit e supplierth kth o t s e risTh f .o k conversion of earnings in local money to internationally convertable currenc n thii s i y case with the buyer.

2.3 BARTER BASIS FOR EXPORT FINANCING An additional risk of the buyer's country is the conversion of raw-materials like crude oil or metalli e worlc th ore dn o smarke t into internationally convertable currency. So the , buyer's country is interested to shift this

492 risk to the plant supplier. He can use the earnings of this barter deal for the reimbursemen e supplier'th f o t s credite Th . risk of unstable monetary values of the raw-materials accordin o variationt g e th n i s world market trading prices then lies with the supplier. The foreign buyer remains responsible for the plant operation after handovee th d an r managemen f energo t y suppl d earningan y s (fig. 7).

~~Raw I materials Official authorities: Supplier L Earnings Ministry of finance/economy Central Bank order Commodities~~

~~Bank (exporter's Credit~~

~~Plant~~

~~20 -MB +16 +14 +12 +10 +8 +6 +4 +2 2 -4 -6 -8 -10 Years~~

~~• Money (internationally convertable) .--- Local money K=K Raw materials, commodities - - —Function (unfillee b o st d Pig: 7 Expor. t Financin n Barteo g r Basis~~

2.4 JOINT VENTURE BASIS FOR EXPORT FINANCING e mosth t f o sever e On e probleme th n i s P projecmanagemene succesNP th a s i f to s f o t commercial operation of the plant. This requires a well-trained and experienced operation staff r thiFo . s rise foreigth k n

493 buye e s experiencinterestei rth e us o f t o de the supplier during the period of commercial operation. This can be arranged by foundation of a joint-venture utility, responsible for power plant operatio d energan n y management. The risk of conversion of the earnings from local currenc n internationalla o t y y convertable currenc s thei y n with this utility. From ther monee th e y flow plane s th bac to t k supplie r reimbursemenfo r e supplier'th f o t s credit (fig . Als8) . mixtura o f joino e t ventur d bartean e r dea s possibli l e (fig. 9) .

~~Raw materials Official authorities: Ministr f finance/economyo y Commodities Central bank~~

Construction I———I———l———I———I———I———I——I———» I———I——I———! 6 - -8-1 4 - 02 - 0 2 + 4 + 6 + 8 + 0 +1 2 +1 4 +1 6 +1 8 +1 0 2 *

~~—— Money (internationally convertable) -_-- Local money •UK Raw materials, commodities -- —Functions to be fulfilled~~

~~Fig. 8: Export Financing on Joint Venture Basis~~

~~494 Official authorities: Ministr l finance/economyo y Central bank~~

~~^200 %~~

~~6 - -8-1 4 - 0 2 20+18+16+14+12+1- + 0 2 + 4 + 6 + 8 + 0~~

~~•Money (internationally convertable) ---- Local money -K=H Raw materials, commodities -- — Functions to be (unfilled~~

~~Fig : 9 Expor. t Financin n Joino g t Venture Basis and Barter~~

2.5 MIXED SOURCE R EXPORSFO T FINANCING In special situations a mixture of various financing sources can reduce the credit amount fro e bane grantee b th m k th o t whico t ds ha h supplie r buyer)(o r . Such a situation occurs when money/ for example, from development aid can be granted by the exporter's country with subject to the implementatio a specia f o n l project a suc s a h nuclear power station.

495 3. CONCLUSIONS Financing of NPP projects becomes easier with HTR plants due to low total investment cost low construction periods good-minded operation characteristics high safety standards and inherent safety high local participatio y proveb n n conventional technologies and smaller dimension e componentth f so s used. It is not possible to discuss the best solutio r financinfo n a genera n i g l mannet bu r there are a lot of suggestions indicating that export financing is possible even when the creditworthines e buyer' e buyeth th r o rf o s country does not exist. All these financing models sho a tendencw y to shift financial risks to the supplier's side. This increases the complexity of the project during implementation. In cas f barteo e r financing riskf o s international trade additionally are shifted to the supplier's side. In the case of joint venture's additionally the ris f operatioo k e planth s shifte i f to n d partiall e supplier'th o t y s side. Froe supplier'th m s poin f vieo t w these increases in financial risks must be compensated by guarantees on the political and financia le buyer' leveth f o ls countrd an y additionally by export credit insurances of the supplier's country for successfull implementation of an HTR nuclear power project.

References; | IAEA-TECDO1 | C 378 Costs and Financing of Nuclear Power Progammes in developing countries Proceeding Siminaa f o s r 09-12 Sept. 1985 | 2 | Brandes, Kohl, Schmitt Small Nuclear Power Plants: R Gas-CooleGH W M 0 1 d Heating 0 Reacto10 d an r MWe Industrial Nuclear Power Plant IAEA Technical Committee Meeting on Gas-Cooled Reactors and their Application 20-23 Oct. 1986, Jülich, F. R. of Germany

496 3 l E. Baust, J. Schöning HTR-500: The Basic Design for Commercial HTR-Power Stations IAEA Technical Committee Meeting on Gas-Cooled Reactor d Theian s r Application 20-23 Oct. 1986, f Germanjulieno . R . yF , . Stösse. R al . et l| 4 Finanzierungsformen im Exportgeschäft in E. Dichtl . Issing,0 , Editor f "Exporto s s al e Herausforderun e Deutschdi r fü g e Wirtschaft" Deutscher Institutsverlag 1984, Sachbuchreihe , ISB38 N 3-602-34842-3

Lis Abbreviationof t s = Brown C G Mannhei,BB A Bovere Ci & mi = HeavHM y Metal HRB = Hochtemperatur Reaktorbau GmbH Mannheim/Düsseldorf = Hig hR TemperaturHT e Reactor PWR = Pressurised Water Reactor DM = Deutschmark MWd/t = Megawatt-Days per ton (of HM) NHSS = Nuclear Heat Supply System NPP = Nuclear Power Plant

~~497 PERSPECTIVE HTGE TH RN FROSO M A UTILITIEUS E TH N SI~~

~~H.L. BREY Public Service Compan f Coloradoyo , Denver, Colorado L.D. MEARS Gas-Cooled Reactors Associates, San Diego, California United States of America~~

~~Abstract~~

s CooleGa e d Th Reactor Associates (GCRA) represents utility/user interests within the United States in establishing and achieving HTGR program objectives. GCRA conducts Demonstration Project development activities, plus provides overall program coordination/integration support services . S Departmen . U e th f Energyo to t . This paper summarizese th thre f o e activities currently being undertaken by GCRA. These include:

~~0 The 1936 Summary of the United States Utility/User Questionnaires concerning new electrical generating capacity, the future of nuclear e potentiapower th e HTGR d th ,an f , o l~~

~~0 The utility/user requirements for the modular HTGR plant, and~~

~~0 An update on the modular HTGR Demonstration Project development activities.~~

Introduction This paper is a review of three significant programs currently being undertaken within the U.S.A. by the Gas-Cooled Reactor Associates (GCRA). These programs include: a summary of the responses received from U.S. electric utility and process heat users to questionnaires concerning new generating capacity e futurth , f nucleao e re potentiath power d f an ,o l the HTGR e utility/usea revie;th f o w r requirement e Modulath r fo rs High Temperature Gas-Cooled Reactor (MHTGR) and; an update on the Modular Gas Reactor (MGR) Demonstration Project. SUMMARY REPORT ON THE UTILITY/USER QUESTIONNAIRE The Gas Cooled Reactor Associates recently compiled the results of a questionnaire 0 respondeutilitie8 3 industriay 1 b d o t an ds l companies as of January 31, 1906. The utility response represents approximately

499 59% of the installed generating capacity and over 72% of the installed nuclear capacity withi e Uniteth n d States e e rankinassociateTh th . n o g d graph a scalo fiv t n f froo so e on mcorrespond e followinth o t s g criteria: Value Interest, Significance r Importanco , e 1 No 2 Low 3 Moderate 4 High Extremely High The significant highlights of the 1985 GCRA utility questionnaire are provided below (Ref. l): Although new power plant orders are scarce in today's environment, the number of utilities planning one or more capacity additions in the late 1990s and beyond is significant. Planned Numbe f Utilitieo r s Inservice Date Planning Additions Before 1995 9 o 199t 95 38 200 200o 0t 4 37 2005 to 2010 29 0 The preferred size of new capacity additions or replacements was e rangeMW 0 s 70 o t 0 40 e th d an 0 divide40 o t d 0 betwee20 e th n (Figur . Thes1) e e result e consistenar s t wite 198th h3 GCRA survey results on size preference, and supports both the reference plant size of 550 MWe (4 modules) and the 278 MWe (2 module) variant.

~~UJ W•z. o ÛL uV) u. o o: U) m 2 !D~~

~~O 2O < 2OO-4OO 4OO-700 7OO-1000 > 1000~~

~~PUVNT SIZE PREFERENCE (MWe)~~

~~Figure I Nuclear Plant Size Preference~~

500 - 200 to 400 MWe favored by small and non-nuclear favoree MW 0 largey b d70 o t r nuclea0 40 - r utilities e averageth n O ,° recommendation n allocatino s g utilit D resourceR& y s were as follows: r maintaininfo % 40 - d improvingan g existing plantd an s - 36% for advanced generation alternatives, including - 11% for advanced nuclear options, including - 3% for the HTGR

~~Obstacles rated highest for new orders for large LWRs (Figure 2): 0 - Financing difficulties - Investment protection concernsd ,an - PUC and NRC regulatory risks~~

~~o 2 n i11 1 111111111111 1~~

~~NEED FMANC C PU E SITINTECHNOLOGG C NR $ LOSPUBLIY S C~~

~~OBSTACLE Figur 2 e Q4-Obstacle Largo t sOrderR eLW s (Averaged)~~

The major obstacles identified that impede the commercialization of the HTGR were (Figure 3): - Difficulties in establishing a combined utility/government/vendor development and demonstration program and - Lack of a strong, established HTGR vendor/supplier infrastructure. Advanced generation options wite highesth h t interest (Figur: 4) e - Advanced coal technologies - Conservation and load management Advanced nuclear generation alternatives received, on average, low o moderatt e interes e HTGtth R wite d improveth indicatinan h R LW d g the highest relative interest (Figur. 5) e

~~501 o 2~~

~~NEED GOVERNMENT UTILITY VENDOR COMBINED TECHNOLOGY~~

~~OOSTACLE~~

~~Figure 3 Q5-HTGR Commercialization Obstacles (Averaged)~~

~~COMPANY PERSPECTIVE AVERAGED~~

~~o z~~

~~COAL SYNFUEL PASSIVE LOAD MGMT~~

~~NON-NUCLEAR GENERATION OPTIONS~~

~~Figure 4 Q6A-Advanced Generation Alternatives~~

~~Except for load management and conservation, company interest in advanced generation alternatives ranked belo e perceiveth w d industry interest.~~

~~502 COMPANY PERSPECTIVE AVERAGED~~

~~o z z~~

~~LWR ALWR HTGR LMR FUSION~~

~~NUCLEAR GENERATION OPTONS~~

~~Figur 5 e Q6A-Advanced Generation Alternatives~~

e preferreTh d institutional arrangemen o ensurt t e succesth e f o s an HTGR Demonstration Project favore e balanceth d d utility, government and industry contro d supporan l t slightly ove a r totaE ProjecDO l t and a total utility-led Project, respectively. 6CRA utilities showed a definite preference for a Project controlled by utilities with complementary support froe governmenth m e vendor/supplieth d an t r industry. The demonstration project objectives receiving the highest ratings were demonstratio d confirmatioan n : of n - Safety characteristics to support design certification, - Overall plant performance, - System and component reliability, - Investment protection characteristics, and - Vendor/supplier capability 0 In selecting between coal, LWR, and MHTGR baseload options that were assumed to be available with equivalent commercial infrastructures, the following results were obtained: - Utilities favored coal over nuclear even thoug e s coath wa hl presumed to have a higher busbar cost by 10-20%; - Coal options, on average, received moderate to high interest and nuclear received low to no interest (Figure 6): - Non-nuclear utilities clearly favored the 300 MWe coal plant and showed no interest in the 600 and 1200 MWe LWR, no to low interest e MHTGMW R0 28 plantw interes e e lo HTGR th MW d ; n an ,0 i t 56 e ith n and

~~503 o 5 z~~

~~R LW O 6O R LW O 2O 1 6OCOAO O3O LCOA L 56O HTGR HTGO 28 R~~

~~BASELOAD PLANT OPTION~~

~~Figur 6 e QlO-Interes Baseloan i t d Alternatives (Averaged)~~

Nuclear utilities showed moderate to high interest in the coal options, and relatively low interest in the MHTGR and LWR options, although their interest in the nuclear options was much greater than for non-nuclear utilities. GCRA utilties clearly preferred HTGR options ovey othean r r nuclear alternative. - Utility preferences for the advanced reactor systems was highest foTorr thee improved LWRK ui- wite HTGith ha R the HlbK a closclosee secondsecond.. Thmee HTGHibRK was ranked higher than either the advanced LWRs (e.g. PIUS) or the LMRs from both the company and industry perspectives. The significant highlight of the 1985 GCRA industrial questionnaire responses is provided below:

Process heat users are looking to utility cogeneration and coal 0 for future energy supply with the least interest in nuclear or externally generated electric energy sources (Figure 7). UTILITY/USER REQUIREMENTS FOR THE MODULAR HTGR PLANT A major role of GCRA is to establish utility/user requirements for the MHTGR design. A summary (Figure 8) of these requirements for a MHTGR electric generating plant with 4 reactors and 2 turbine generators with an electrica t includesne e MW l :0 outpu55 f o t

AVAILABILITY e equivalenTh t availability average ovee lifetime th planrth tf o e r electricafo t leasa shal% e 80 b t ll generation when modeled using e equipmenth t mean tim o failuret e , mean tim o repait e r datr likfo a e type, or similar systems and/or components.

~~504 o £~~

~~D. z a o~~

~~EL EC SYN COAL NUC COG EN OTHER~~

~~TYPE~~

~~Figur 7 e Process Hea tAdvance- Usel Q - r d Energy Preference (Averaged)~~

CRITERIA UTILITY/USER REQUIREMENT ANNUAL AVAILABILITY TOTAL OUTAGE 20% MAXIMUM OVER LIFETIME SCHEDULED OUTAGE 10% MAXIMUM OVER LIFETIME PLANT INVESTMENT PROTECTION UNSCHEDULED OUTAGE 10% MAXIMUM OVER LIFETIME OUTAGE MONTH6 > S S 10% MAXIMUM OF UNSCHEDULED OUTAGES EXPECTED VALUE or LOSS < ANNUAL INSURANCE PREMIUM PROBABILITY OF EXCEEDING SAFETY RELATED DESIGN LIMIT 10'5/PLANT-YR SAFET LICENSIND YAN G OVERALL CRITERIA EXISTING NRC/EPA DOSD AN E RISK CRITERIA EMERGENCY PLANNING CRITERIA NO SHELTERIN R EVACUATIOO G N REQUIRED SITING PARAMETERS EXCLUSION AREA BOUNDARY RADIUS 425 METERS SEISMIC (GROUND ACCELERATION) .SSE/.1g 3 E QB 5g L CYCLE ENRICHMENT LEVEL % 20 LOW < , SPENT FUEL MANAGEMENT ONCE-THROUGH THROWAWAY

~~ECONOMIC GOALS BUSBAR POWER COST AT LEAST 10% ADVANTAGE OVER COMPARABLY SIZED ADVANCED COAL PLANTS INSTALLED CAPITAL COST < 20005/KW (1986 DOLLARS)~~

~~Figur e8 Summar f Utility/Useo y r Requirement r MHTGfo s R~~

505 e equivalenTh t unavailability averaged ovee lifetim e th planr th f to e because of scheduled outages shall not exceed 10%. (This is an average 6 hrs oyear)r 87 f .pe . PLANT INVESTMENT PROTECTION It is a design goal that the unavailability averaged over the lifetime of the plant because of forced outages shall not exceed 10%. Outages which last 6 months or more shall not contribute more than 10% of the total unavailability from forced outages. The calculated financial annual value for risk to plant equipment r properto y when averaged ovee planth r o texceet lifetimt dno s i e e annuath l property damage insurance premium usen i economid c assessments e currenTh . t assumed e portioannuath f lo n premiur fo m property damage is 4.5 million dollars U.S. The likelihood of exceeding a design limit associated with safety-related design conditions for a reactor shall be less than 10~r planpe 5 t year. SAFET D LICENSINAN Y G Desig d licensinan n g documentation wile developeb l d wite goath h l f obtainino g design certificatio e standarth f o n d Nuclear Islany b d the NRC. This includes meeting existing NRC/EPA dose and risk criteria. This plant shall be designed to meet regulatory requirements without taking credit for the need of sheltering or evacuation of the public beyond the plant's exclusion area boundary. SITING PARAMETERS The radius of the exclusion area boundary shall be 425 meters. The horizontal ground acceleration seismic criteria will be .3G for the safe shutdown earthquak e d operatin.15th an e r 6fo g basis earthquake. FUEL CYCLE This modular reactor plant will hav a fuee l cycle basee dus upoe th n f o once-throug w enrichelo h d uranium. However, other fuel cycles f higo e h us enriche e th suc s da h fue r recycleo l d fuele b whic n ca h utilized without requiring additional modification to the basic core/fue e lphysica th design o t lr o , plant o refuelint r o , g components will be identified by the completion of this conceptual design effort.

ECONOMIC GOALS A design goaf thio l s o havt plan n economia s ei t c advantagt a f o e least 10% in the 30 year busbar cost of electricity relative to a comparable sized state-of-the-art coal plant. Based on current coal plant estimates, this results in a target of approximately 45 mils per KW total busbar cost and a total capital cost of approximately $200 r installe0pe W electriK d c including contingencies, owners' costs, and AFUDC (Ref. 2).

~~MODULAR HTGR DEMONSTRATION OBJECTIVES As indicated previously, GCRA, under the direction of its members, s undertakeha n development activitie e eventuath r fo ls constructiof o n~~

506 a modular HTGR which woul s purposd it hav s o demonstrata et e e followinth e g objectives (Figur: 9) e 0 Demonstrate licensing process and support design certification Through the Reference Plant development effort, a disciplined requirements-based approac o licensint h s beini g g developed that should fully capitaliz e MGR'th n so e unique safety characteristics. This process will be put to practice and demonstrated through such a Project. 0 Demonstrate Plant Performance Performance of a full-scale reactor module and its interaction withi e Nucleath n e r overalth Islan d an l d plant wile b l demonstrated. In addition to the normal steady state power operation and anticipated transients demonstrations, design basis events may be demonstrated to verify inherent safety and investment protection feature e plantth f .o s o Demonstrate Maintenanc d Reliabilitan e y e ProjecTh t will provida demonstratio r fo e f normao n l operation and maintenance (O&M) activitie s wel a ss identif a l d demonstratan y e O&M which might be required in the unlikely event of major equipment failures. In addition, a reliability improvement program will be establishe n integraa s a d le Project th par f o t . o Establish basis for commercial plant cost and schedule e projecTh t will provide first-of-a-kind experience with adapting modular fabrication techniques and separation of the Nuclear Island

~~DEMONSTRATE THE LICENSING PROCESS USING THE CRITERIA AND METHODOLOGY ESTABLISHED FOR THE MHTGR PROGRAM AND SUPPORT DESIGN CERTIFICATION EFFORT S REQUIRED A E STANDAR, TH R FO ,D MHTGR~~

~~DEMONSTRATE PLANT PERFORMANCE CHARACTERISTICS~~

~~DEMONSTRATE CRITICAL MAINTENANCE ACTIVITIES AND SYSTEM/COMPONENT RELIABILITY~~

~~ESTABLISH BASIS FOR COMMERCIAL PLANT COST AND SCHEDULE PLUS FOSTER THE DEVELOPMENT OF A VENDOR/SUPPLIER INFRASTRUCTURE~~

~~ESTABLISH UTILITY/USER/INVESTOR CONFIDENCY BU O T E COMMERCIAL PLANTS~~

~~Figure 9 MHTGR Demonstration Project Objectives~~

507 and Turbine Island portion e plant th e cosd schedulTh f an .to s e experienc e overalth n i le design, licensing, fabricatiod an n construction effort will serve as invaluable demonstration experience for prospective vendors and customers. 0 Develop Vendor/Supplier infrastructure e ProjecTh t will foster o e mordevelopmenth ee r on f o t vendor/supplier entities and related infrastructure for offering subsequent commercial plants. 0 Establish Utility/Investor confidence Once the objectives above have been met, the utilities and the financial community would have sufficient confidence to proceed with commercial units (Reference 3). As an initial requirement of this modular HT6R demonstration project, a definition stud s undertakewa y n beginnin n Augusti g , 1985 e overal.Th l objective of this study was to provide an assessment for determining the feasibility of a M6R Demonstration Project. The tasks associated with this study includ e developmenth e e demonstratioth f o t n plant scopd an e layout, establishmen e projecth f o t licensing approach, evaluatiof o n utility siting options, developmen a preliminar f o t y testing program, definition of the supporting technology requirements, an estimate of total program a costsdefine d an , d program schedule (Figure 10). Figur1 1 e provides the estimated' time frame for the phase-in of the modular demonstration plant.

~~• SCOPS 6 LAYOUT • LICENSING APPROACH • SITING OPTIONS • TESTING PROGRAM • TECHNOLOGY REQUIREMENTS~~

~~« PROGRAM COST • PROGRAM SCHEDULE~~

~~Figur 0 Projec1 e t Definition Study Objectives~~

Summary Utilities representing approximately one-third of the installed generation capacite Uniteth n di y States for a majom e rth parf o t constituenc f GCRAr o utilityOu . y membershi a stron s ha pg interesn i t seein e developmenth g a future HTG s th a R f eo t major electrical generation e Uniteoptioth n i dn e Statescommitte ar e continuin e th W .o t d g development of nuclear power with specific emphasis on the HTGR. The significant attributes provided by this advanced concept will play an important role n energa s a y e futuresourcth f o e.

~~508 PHASE I - PROJECT DEFINITION (THRU 1988)~~

~~- PHASPRELIMINAR A II E Y DESIGN AND LICENSING (1989*1990)~~

~~PHASE IIB - DETAILED DESIGN AND LICENSING/LONG-LEAD MANUFACTURING (1991-1992)~~

~~PHASE III - MANUFACTURING, CONSTRUCTION AND TESTING (199 31997- )~~

~~- PHASCOMMERCIALIZATIO V I E N~~

~~Figure 11 MGR Demonstration Project Phases~~

~~REFERENCES~~

~~1. Gas-Cooled Reactor Associates, Summary Report on the Utility/User Questionnaires, GCRA 86-004, February 1986.~~

~~. 2 Gas-Cooled Reactor Associates, Utility/User Requiremente th r fo s Modular High Temperature Gas-Cooled Reactor Plant, GCRA 86-002, Rev. 2, 1986.~~

~~. 3 Gas-Cooled Reactor Associates, Modular Gas-Cooled Reactorsn A , Attractive Future Electric Generation Option, Designed to Meet Utility Needs.~~

~~509 MARKET PROSPECTS OF HTRs IN NEWLY INDUSTRIALIZED AND DEVELOPING COUNTRIES~~

~~. GARRIBS A Politecnic i Milanood , Milan, Italy C. VIVANTE Commission of the European Communities, Brussels~~

~~Abstract~~

MARKET PROSPECTS OF HTRs IN NEWLY INDUSTRIALIZED AND DEVELOPING COUNTRIES Some unique characteristics seem to make HTR a nuclear system suitable to cope with future powe d energan r y supply need f developino s g world. Aspects which deserve consideration are economics, technological flexibility, safety and proliferation issues. As compared with other nuclear reactor concepts, HTRs can be constructed into relatively small sizes according to a standardized approach. This characteristic renders the reactor suitable to be introduced in power grids of limited extension e smalTh . l size allow o increast s e installeth e d power generating capacity by following the variable trend of demand, all within a short construction period. The high degree of inherent safety makes the reactor system especially resistan o equipment t t failure d mismanagementan s . Safet d high-temperaturan y e feature e reactosth woulr f o fo r de us allo e th w supplying stea d procesan m s hea a newlt y industrializing countr y needma y n I . principle, non-proliferation record f HTRo s s depend upoe typ f th nucleano e r fued nucleaan l r fuel cycle f low-enricheI . r medium-enricheo d d uranius i m used, reprocessin gt necessary woulno e b d e exhausteTh . e easildb fuen yca l accomodate d storean d d before ultimate disposal s relatio. it Finally r fo n s ,a with the local context, HTR entails flexible arrangements, which can be adapted to participation and seem suitable to technology transfer. Given these opportunities, it can be suggested that forms of co-operation with developing countries are looked for to monitor technology progress and prepare appropriate solutions.

. PREMIS1 E e newlTh y industrialize d developinan d g countries represent a wide spectrum of states and their interests towards nuclear commitment vary substantially. Some nations are economically and industrially better placed. Few ones are exporters of energy, while most of them are in a shortage of indigenous energy resource d technologiesan s A .widesprea d reactioe th o t n basic issues of energy is that the international community has shown very little car r intereso e understandinn i t d helpinan g g resolv e acutth e e energy problem e Thirth df o sWorld , although these problems are paralyzing national economies and threate- ning their survival e recenth n I t. past e risinth , g costf o s e growinth d oian gl uncertaint s supplit n i y have undermined

511 growth perspectives and created balance-of-payment difficulties. Although the price of oil has now fallen in real terms, countries in the Third World remain concerned about their energy supplie d considean s r nuclea n economia r e poweb o ct r and viable alternativ r meetinfo e g their power needs. Instea f relyin o dn earl a n yo g adven f sofo tr exotio t c technologie r quicklfo s y rescuing from energy shortages, quite w developinfe a g countries continu o believt e e that serious attention should be paid toward making nuclear power econo- micall d technicallan y y more accessibl r extendefo e d use, under appropriate safet d safeguardan y s regimes. Two main general difficulties are limiting an effective development of the nuclear option, however. First, are the delays accumulated by nuclear programs in the major industrial countries. Second, are the serious problems encountered in nuclear technology transfer and adaptation. Needless to say that nuclear programs of the major industrial countries are at a standstill. This is partly e becausslowinth f o eg dow f nationao n l energy demand growth projections d partlan , y becaus f o strone g oppositiod an n pressures from certain groups that express fears aboue th t environmental impac f nucleao t r power. Therefore some observers preten de technologicall th tha f i t y more advanced natione ar s curtailing their nuclear power programs, there is perhaps even less justification for developing countries to turn to nuclear energy. This argument does not allow for the fact that the sustained increase in power supply is a prerequisite for economi d industriaan c l developmen e othee sideth on n rn O .o t side, use of nuclear energy would ease the problem of proper managemen e naturath f o tl resource base. Suc a hpolic y goas i l especially importan n developini t g countries, because these countrie n leasca s t affor e opportunitth d y cost f irreversio s - ble los f theio s r renewable natural resources r efforto , o t s remedy environmental damage. The worry tha e spreath t f nucleao d r technolog y acceyma - lerat e proliferatioth e f o nuclean r weapons adds further concerns s beeha nt I .sometime s argued that newly industrialized countries have sought nuclear reactors, not out of any serious interes n commerciai t l power generation t rathebu , o t r enhance national prestige o gai,t n acces o nucleat s r technology e antechnologicath d n trainingo t ge o lt , trajector f nucleao y r learning. Some countries indeed seeo t havm e based their decision o start s p theiu t r nuclear program n sucso h considerations n thiI . s process, companies from industrialized economies were sometimes led to exaggerate the promise of nuclear power and to understate its risk. Developing countries were advised, for example, to prepare themselves by installing large power o reactort tak d e an sadvantag e from scale factorsa s A . result f nucleai , r reactor e majoo provb d s o rt e technicar o l financial burden (and this likelihood cannot be ruled put), a frustrated developing country has no means of justifying the substantial investment in civilian reactors and human resources. r thifo ss A special difficulty related withn a lac f o k adequate industrial and technological base in developing countries, a castherr designins fo ei e d constructinan g g standardized, small or medium-sized power reactors that could be operated in needy regions, and technically serviced through

512 international arrangements to ensure proper and efficient operation [l,2Jw patne h . A shoul e founb d d toward e developth s - ment of uses of nuclear energy in the Third World. HTRs basing on small sizes and highly standardized units, like the ones proposed by some major nuclear vendors, seem to offer a viable option [3-5]. In this frame, four main problem areas seem worth f evaluatioo y d wil e an considerenb l e followingth n i d , economics, role of local participation, safety, and proliferation characteristics.

E ECONOMICTH . F 2 HTRO S s e externaTh l cos f constructino t d operatinan g g nuclear power plant d theian s r facilitie n developini s g countriea s i s key and still unresolved problem. This fact is not surprising, since few developing countries possess the infrastructure which is required for extensive participation in the construction of nuclear power plants ordered from foreign vendors. Usually, developing economies have negotiated turn-key packages that include arrangements for the financing of the nuclear projects y foreigb n vendors s plana .r tWhilfa operatio s a e s conceri n - ned, most of the commercial nuclear power stations in the developing countries have come on-line recently d theran ,s i e insufficient experienc e e runnininstallationsth th n f i eo g . Thus real costs over time canno e assesseb t d wity degrean h f o e accuracy. The question of the costs of nuclear power is, of course, not peculiar to developing countries. Industrialized economies confront themselves e witsamth he issue, although wita h different set of constraints and a different framework of comparative choices. Wha s speciai t o developint l g countries i s their neeo considet d r additional decision issues thaty b , definition, appear to be settled for advanced countries. Among these special issues are: e structur Th e nationa th ) (i f o e le th grid s sizd it ,an e portion that a nuclear power plant would contribute to it; (ii) The search for and the access to financial means from indigenous and external sources to support the project; (iii)The degree of industrial and technological maturity of the country. Let us examine these three issues one by one. Regarding unit size, it is apparent that present-generation commercial reactors, like LWRs, only operate economically abov a certaie n minimum threshold of installated capacity. Such a break-even is in practice dictated in advance by the fact that, with present economie f scale o sonls i yt i , those reactors whic e workinar h g in large numbers already whic n competca h e with other sources of power. These reactors are currently located in the industrialized countries and their size exercises a determining influence on the reactor market through the world. Conversely, in almost any developing country, even a single reactor of the minimum unit size now readily available in international trade would represen a tlarg e proportiof o n the total electricity supply system, with obvious implications for the vulnerability of the same system to the withdrawal of a single generating unit from service. Technical considerations dictate that only a limited percentage of the total electricity

513 outpu a countr f o t y shoul e generateb d e individuaon y b d l power station (regardless of how the unit is powered). Would this limit be exceeded, the danger exists that any unplanned outage in the station could create problems throughout the system, and maybe even o breacaust t i ke down totally. This also applies where there is no national transmission system, and where a single power statio s responsibli n r supplyinfo e a particulag r economic or demographic area: in such a case the risk exists of a black out. In principle e designe,b HTRn d ca constructesan d o t d overcome these problems. Adoptio f smalo n l sizes would alloo t w follow an approach of standardized design and construction, and o avoit d overcosts connected wite diminishinth h ge scalth f o e plant A .considerabl e eplan th par f to t coul e shop-construcb d - ted, rather than erected in the field. Concerning the issue of financial resources, two key limiting factors are obvious: first, the level of real interest rates d secondan ; e availabilitth , f externao y l capital. There are other important factors too, chiefly the allocation policy f scarco e national investments e remarkeb n ca dt I .indee d that the real cos f nucleao t r electricity f creatino , d supporan g - ting a nuclear generating capacity, with the necessary industrial and regulatory infrastructure, may preempt a larger share of the capital available for investment in energy supply systems than does a generating systems designed for fossil fuels. Again small standardized HTRs would ease these constraints. By allowing a gradual approach to nuclear power development it would become possible to plan a balanced distributio f fundo n s whic y becomma h e accessible. Moreover, HTRs could be adopted to produce electricity, or process heat, or both. This means that the same reactor can cope with the variet f electricao y d othean l r energy supply requirementn i s the country, so that possible uses of nuclear technology appear extended [6]. . 3 ENERGY SECURIT D INDUSTRIAAN Y L DEVELOPMENT The issue of the degree of industrial and technological maturity which is needed to embark on a nuclear power program, deserves careful examination. In a developing or newly industrialized country, the introduction of nuclear power may help to diversify supplies of energy in general and electricity in particular. The dependence on any one source of supply would be diminished and the dependence on imported energy sources would be reduced correspondingly. However, nuclear electricity demands a rather elaborate apparatus, in the shape of a reactor and its safety features, and a considerable series of industrial processes to transform uranium ore into the fuel for such a reactor. The most difficult questions about energy security in connection with nuclear power arise, not from comparing dependence on uranium to that on fossil fuels, but from considering what other forms of external supply a developing country must rely upon in launching a nuclear power program. A common dilemm s beeha an that technology imported from industrialized countrie s oftei s n difficul o adapt t o locat t l socio-economic conditions. Such technology transfer can sometimes even hamper scientific innovations and entrepreneurial initiatives for indigenous development, leading to repeated

514 technology imports. Yet many developing countries do not have a cadr f scientifio e d technicaan c l personnel e capablar o f ,wh o e advising their national authorities on a strategy for the introductio a sophisticate f o n d technology, suc s nucleaa h r energy, which would be suitable to local conditions. Most countries do not have adequate resources for establishing their D effortsw developinR& fe n e ow Th . g countries that have th e financial resources are, by and large, short of manpower and without adequate educational infrastructure. Another constraint whic developinl al h g countries have absenc o th fact a es i ef o e diversified industrial base. Muce sam s th truh i e f dependenc o e n externao e l nuclear fuel-cycle e servicescas th f reacto o en i s A .r manufacturer ,fo energy security, there is an obvious reason for seeking to reduce national dependenc n foreigo e n fuel-cycle servicesA . country choosin nucleas git HTRr rfo s program needs accesst ,no onl o naturat y le serviceth uranium l d al an t sals o bu ,t o processes involved in enriching uranium and fabricating suitable fuel elements. There is a question of economical, as well as technological, prudence in relation to indigenous fuel cycle development, however. Especially, careful judgment is required with e establishmenregarth o t d f enrichmeno t r reprocessino t g capacit a singl n i y e country. These limitations would not prevent that benefits of a well-planne d successfullan d y executed nuclear prograe ar m industrial and technical, as well as energy-related, because the program can result in the progressive accumulation of knowledg d formatioan e f constructioo n d maintenancan n e apti- tudes. It is admitted indeed that HTRs entail a flexible technology, which can be adapted to a variety of local situation d prerequisitesan s e desigth o t n e characteristicDu . f o s the reactor system, opportunitie r locafo s l participation i n building HTRs could and should be exploited in ways that promote a generally higher level of technical and industrial expertise, embodie a skille n i d d work forc d deployean e th n i d for f manufacturino m g capability e consequenceTh . s woula e b d lasting benefit to the economy concerned. These prospects must, therefore, be in the planners' minds when they asses e impacth s f embarkino t e nucleath n o gr program n theiO . r part, nuclear vendors ough o considet t e th r problem of technology transfer as a strategic goal in the developmenR conceptHT e th . f o Accest o advancet s d technology and industrial skills neede a nuclea n i d r prograe seenb y ,ma m ia widen f raisinro context y e levewa th gf scientifi a o l s ,a c and technical development, jus s electrificatioa t n relyinn o g nuclear power may be seen as an optimal path to economic development based on industrialization. Clearly, for technology transfer can succeed, further investments would be required to create industrial capacity, onl a fractioy f whico n h mighe b t anything like fully employed once an initial reactor program has been completed.

4. SAFETY AND ENVIRONMENTAL ISSUES Basing on both experimental and theoretical evidence, smal d medium-sizan l e HTRs exhibit favourable characteristics relativ o safett e d environmentaan y l concerns. Inherend an t

515 passive safety features make these reactors a low risk to the public and to the operating personnel. A difficult matter for the planner to consider is that of e public'th s acceptanc f nucleao e r y powerseema mt I .strang e o quott e thi n issua s r developinfo e g countries, becausr fa o s e it is believed to be in advanced industrial countries alone that serious problems of public opposition to nuclear power have arisen. But aftermaths of Chernobyl disaster have made apparent once more, that an accident which happen in whatsoever nuclear plant has a negative feedback on the overall nuclear market n thiI . s view e higth , h degre f inhereno e t safetf o y HTRs makes these reactors particularly resistant to equipment failure d mismanagementan s . Inherent safety feature f moso s t HTRs relate primarilo t y fundamental law f natureo s . These featuree firsth te ar sline s of defence in maintaining the integrity of the barriers against the release of radioactivity. An additional characteristic of critical importance is that the consequences of accidents develop slowly, allowin e operatorgth timr fo eo tak t s e delibe- ratplanned an e d actions [7]. All aspects of inherent safety features, engineered safety equipment, barriers to fission products have been discussed at lengt y otheb h r authord furthe ad o d nee n thero an t ss d ri e remarks. These aspects seem also more valuable for countries where the lack of adequate skills and resources might lead to operate nuclear reactors along sub-optimal conditiond an s procedures. Particularly, well known passive safety characteristicf o s the HTR are: (i) A negative temperature coefficient of the core; (ii) Coated-particle ceramic fuel that releases fission products slowly, even under extreme temperatures; (iii)A graphite moderator/reflector that responds smoothly to transient temperature, while maintaining structural integrity at very high temperatures; (iv) Helium coolant which remains a single-phase gas under all conditions, and is neutronically transparent and chemically inert. Should accidents be considered, engineered safety equipments that contribute to their low probability and low consequences, include: (i) The triply redundant core auxiliary cooling system (CACS), whic s usei h d when main loop coolin s unavailablei g ; (ii) The reserve shut-down system, which can bring the reactor o coot l shut-dow e absencth n i f ncontroo e d insertionro l ; (iii)The plant protection system, which initiates various safe- y actionst , suc s reactoa h r trip, steam generator isolation and dump, and CACS start-up. The usefulness of probabilistic risk assessment has been demonstrated especially with LWRs. It would be appropriate to extend this typ f analysio e o embodt o HTRss t a sy o externa,s l events and operating conditions which could be encountered in developing nations. Lastly, as for the environmental concerns, it can be recalled the high efficiency of HTRs in the conversion of thermal energy into electrical energy even with dry cooling. This fact means less stringent site constraints because th f o e reduced low-quality thermal waste which mus e dischargedb t .

~~516 Another possibility which has its own merits and does not reserve further comment s cogeneratioi s o copt n e with industrial uses of energy.~~

5. THF SAFEGUARDS PROBLEM Developing and newly industrialized countries are con- e facsciouth t f o thas t prevailing fears about proliferatiof o n nuclear weapons are an obstacle to their acquiring nuclear technolog r peacefufo y l purposes. Nuclear proliferatios i n basically a political problem. Hans ßlix emphasizes that "the first and most important obstacle to the proliferation of nuclear weapon a matte s i sf politica o r l judgmen d deteran t - mination, as it emerges from assessments of political and security conditions f benefito , s possibly accruing froT NR m affiliation, and of drawbacks possibly connected with retention of the nuclear-weapons options" [8J. The reasons for the proliferation of sensitive nuclear technology seem indeed linked with perceptions of security needs and national interests. However, should a developing country decid o acquirt e e sensitive technology while following a progra f nucleao m r power development woult i , d inevitablt pu y an almost unbearable economic burden and cause serious social and political problems within its own borders. Any unbiased analysis would indicate that this would be too high a price to r somfo ey kinpa f nucleao d r military capabilit f doubtfuo y l effectiveness. e meaninfgub y y caseIma an n t o ,analyzi t l e importanth e t proble f o non-proliferatiom n a objectiv n i n e manner with referenc o HTRt e . Proliferation concerns vary widely wite th h type of fresh fuel utilized. All proposed HTR versions and fuel cycle options rely upon different grades of enriched uranium. For the low-enriched uranium and medium- enriched uranium fuel cycles the uranium in the fresh fuel itself would not be weapons usable without further enrichment t shoulI . e alsb d o noted that once the fuel elements are fabricated, the fuel is present in a highly diluted form, due to the coatings and the associated graphite moderator material. Once exhausted, fuel can be easily accomodated and stored before ultimate disposal. And if reprocessing should be maintained as a possibility for e futureth , technology would still need further studd an y experimentation. The different HTRs requir e implementatioth e f safeo n - guards procedures according to their special design features, mainly off-load refuelling for prismatic elements and on-load refuellin e pebble-beth r fo g d version e accesssibilitTh . n a f o y individual fuel elemen s quiti t e different, whethe e eleth r - n fixei ment e d ar spositions r movo , e randomly througe th h core. The safeguards systems take these differences into consideration. Despit e recenth e t high visibilit f terrorismo y , R woul at HT greatln no d y augment terrorist's abilit o threat y - ten damage, even with the help of a co-operative plant insider inten n o causint g core damag r o some e other radiological incident. Compared with other reactor systems, too much emphasis has not to be placed on physical deterrent measures which might not be as effective against insider threats from one side, and which might impair emergency operability from the other.

517 6. CONCLUSIVE REMARKS Most industrialized countries believe that nuclear energy offers a viable solution to their electric power problems. The construction and operation of nuclear reactors, with the concomitant w savingmaterialra n i sd fossi an s l fuels along with the protection of the environment, is essential to continuing economical growth and industrial productivity. There are strong incentives for developing and newly industrialized countries to turn to a technology that is already proven and e generatioth r fo f e o n us o t availablet pu e b d than an ,ca t electricit d possiblan y y heat productio t competitiva n e costs. e motivatioth Thi s i sr integratinfo n g HTR n Thiri s d World economies a flexibl s a , e solutio o yielt n d industrial, techno- logica d energy-relatean l d benefits. While expanding their capabilit r usinfo y g nuclear energy, Third World countries cannot alonei o g t, however. They need meaningful co-operation froe industrializeth m d countried an s an understandin e problemsth f o g . They require sharind an g transfer of technology and dependable supplies of equipment, materials, and services. They ask for help in acquiring appropriate management skills, know-how d financing,an . r assumptioOu s thai n t small-size standardized HTRs have several characteristics that make this reactor concept suitable to cope with the special requirements set forth by developing countries. In the recognition of this evidence, it seems appropriate that potential nuclear suppliers and possible host countries take a number of positive measures aimed at building a market for HTRs, owing to the fact that this reactor concept and syste e closar mo commerciat e l deploymen e morth en i t industrialized economies. Forms of inter-country co-operation can be perhaps devised to collect, evaluate and disseminate a wide range of information about HTRs and their applications, including such question s nucleaa s r power (and heat) economics, safety and environmental considerations, plant performance and operation, all with particular application to the Third World. Formal associations or interest groups might also be created, where potential nuclear suppliers and recipient countries work together to finalize HTR design, construction and operation. Several e actionproposeb n ca s d regarding adaptatio f HTRo n s to special environments, support to licensing procedures and safety analysis, collectio f dato n a concernin e operatioth g f o n existing HTRs, organizatio f o locan l industr d operatoan y r training, establishmen f joino t t ventures. Finally e scopth , e of the work could be broadened to look at nuclear power option in the context of total energy needs and resources. Every available opportunity should be used for conducting a serious dialog between the Third World and the more advanced nuclear countries to comprehend each other's perceptions and to narrow down existing differences. REFERENCES ] [l Smal Mediud an l m Power Reactors: Project Initiation Study. Phas IAEA, I e , Rep. TECDOC-347 (1985). [2] Small and Medium Power Reactors 1985; IAEA, Rep. TECDOC- 376 (1986).

518 ] [3 BAUST , "HTR-500E. , A .powe r e Statiofuturth r f fo eo n electricit d procesan y s steam generation" d "HTR-100,an .A power station for industry" (Proc. GCRA's 8th Annual Conf. on the HTGR), Gas-Cooled Reactor Associates, San Diego 1 r (1986). [4j WEISBRODT, I.,"Survey of the design status of Intera- tom/KWU's HTR-Module" (Proc. GCRA' h Annua8t s l Confn o . the HTGR), Gas-Cooled Reactor Associates, San Diego . . (1986). [5] Modular HTGR Demonstration Project Definition Study. Summar d Assessmenan y t Report, Gas-Cooled Reactor Associa- tes, San Diego, Rep. GCRA 86-010 Draft (1986). [6] GARRIBBA, S., VIVANTE, C. "Problems concerned with the development of small and medium-size nuclear reactors", e Pane(Procth lf o . Sessio n Safeto n y Issue n Smali sd an l Medium-Size Nuclear Reactor, 8th SMIRT), Commission of the European Communities, Brussels, Rep. EUR, to appear , r (1987). [7J Gas-Cooled Reactor Safet d Accidena y t Analysis (Procf o . a Specialists' Meeting on Safety and Accident Analysis for Gas-Cooled Reactors), IAEA, Rep. TECDOC-358 (1985). [8] BLIX, H., IAEA Bulletin 2]_, 2 (1985)3.

~~519 CHAIRMEN AND SECRETARIAT~~

~~GENERAL CHAIRMAN, VICE-CHAIRMAN AND CHAIRMEN OF SESSIONS~~

~~General Chairman . SchulteR : . Mr n Federal Republic of Germany Vice-chairman: Baithese. E . Mr n Federal Republic of Germany~~

~~Chairmen of Sessions~~

~~Session A Mr. E. Baithesen Federal Republi f Germano c y~~

~~Pexto. A . nMr Sessio B n United Kingdo f Greao m t Britain and Northern Ireland~~

~~Session C Mr. A. Millunzi United State Americf so a~~

~~Session D Vivant. C . Mr e Commissio Europeae th f o n n Communities (CEC)~~

~~SessionE Mr. V.N. Grebennik Unio f Sovieo n t Socialist Republics~~

~~Session F . HayashT . Mr i Japan~~

~~Session G Mr. Schulte.R n Federal Republi Germanf o c y~~

~~SECRETARIAT~~

~~Scientific Secretary: J. Kupitz Divisio Nucleaf o n r Power, IAEA~~

~~521 LIS PARTICIPANTF TO S~~

~~BELGIUM~~

~~Mr. J. PLANQUART Director - Nuclear Program C.E.N./S.C.K. Mol Boeretan0 20 g B-240l Mo 0 Tel. 0032/14/316871 311801 Telex 31922 ATOMOL B~~

~~CHINA~~

~~Mr. Gong YUNFENG Deputy Chief Engineer of Beijing Institute f Nucleao r Engineering P.O. Box 840 Beijing~~

~~FRANCE~~

~~Mr. Y.M.F. BERTHION CEA/CE e Saclad N y DEDR/DEMT B.P. 2 F-91191 Gif-sur-Yvette Tel. 908.24.04 Tele 0 64169 x F ENERGAT-SACLAY~~

~~GERMAN DEMOCRATIC REPUBLIC~~

~~Mr. H.F. BRINCKMANN Zentralinstitu r Kernforschunfü t g Rossendorf, PSF 19 DDR-8051 Dresden Tele k rossendorzf x 7 16 2 0 f~~

~~FEDERAL REPUBLI GERMANF CO Y~~

~~Mr. M. ANDLER Interatom GmbH. Dep. PF113 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

~~523 Mr. G. ALTES Kernforschungsanlage Julien GmbH. IRE Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-0 Telex 83355D A 6KF~~

~~Mr. E. ARNDT Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straß8 e D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. L.T. AUMUELLER NUKEM Rodenbacher Chaussee Postfach 110080 D-6450 Hanau 11 Teled x k 418nu 0 478~~

~~Mr. K.P. BACHUS Bundesminister für Umwelt, Naturschutz und Reaktorsicherheit Husarenstraße 30 D-5300 Bonn~~

~~Mr. G. BALLENSIEVEN Kernforschungsanalge Jülich GmbH. IRE Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 83355D A 6KF~~

~~BALTHESE. E . Mr N Kernforschungsanlage Jülich GmbH. ProjektträgeR rHT Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-554 Telex 83355D A 6KF~~

~~BARNER. H . Mr T Kernforschungsanlage Jülich GmbH. IRE Postfach 1913 D-5170 Jülich Tel. 06-02461.61-0 Telex 83355D A KF 6~~

~~BAUS. E . TMr Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße8 D-6800 Mannheiml Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~524 Mr. B. BEINE Siempelkarap Gießerei GmbH + Co. Abteilung Reaktortechnik Siempelkampstraße 45 Postfach 2570 D-4150 Krefeld-1 Telex 2151338~~

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~~Mr. J. BOGEN BrownG ,A Bovere Ci & i Nuclear Energy Dept. Kallstadter Straße l Postfach 351 D-6800 Mannheim 31 7 10 1 Tele41 2 x46~~

~~Mr. J. BOOSTEN Kraftwerk UnioG nA Departmen 11/RtV l K Hammerbacherstraß4 e1 D-8520 Erlangen Tel. 09131-18-4141 Telex 62929-0~~

~~Mr. T. BORNSCHEIN Kernforschungsanlage Jülich GmbH. HBK Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~BRANDE. S . Mr S Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße8 D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~BRINKMAN. D . Mr N Kraftwerk Union AG Dept9 4 .V Hammerbachstraß4 1 + 2 1 e D-8520 Erlangen Tel. 09131-18-0 Telex 62929-0~~

BRUNNE. H . Mr R NUKEM GmbH. Rodenbacher Chausse6 e Postfach 110080 D-6450 Hana1 1 u Teled xk 418nu 0 478 525 Mr. R. CANDELl Interatom GmbH. Dep. PF113 Friedrich-Ebert-Straße D-5060 Bergisch Gladbac1 h Tel. 02204/84-0 Telex 8878 457 iagl (d)

~~Mr. H.-P. CLASSEN Interatom GmbH. Dep. PF114 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 887 7 iag) 845 (d l~~

~~DECKEr . Cl.-Bde Mr n N.vo Kernforschungsanlage Jülich GmbH. Institut für Reaktorbauelemente Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-0 Telex 83355D A KF 6~~

~~Mr. K. DUMM Interatom GmbH. Dep. PF22 Friedrich-Ebert-Straße D-5060 Bergisch Gladbacl h Tel. 02204/84-0 Telex 887 7 iag) 845 (d l~~

~~Mr. ELTEC . R Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheil m Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. EPPIN.Ch G University of Duisburg Energy Technology Lotharstraße Duisburg~~

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~~Mr. J.A. FAßBENDER Kernforschungsanlage Jülich GmbH. PTH P.O191x Bo . 3 D-5170 Jülich Tel. 06-02461.61-0 Telex 462041~~

~~526 Mr. U. FRICKE Rheinisch-Westfälischer Technischer Ueberwachungs Verein e.V., Steuben Str. 53 D-4300 Essen~~

~~Hr. H. GERWIN Kernforschungsanlage Jülich GmbH. Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~Mr. GLAHE Director Hochtemperature-Kernkraft GmbH Siegenbeckstraße 10 D-4700 Hamm~~

~~Mr. GREM.0 M Kraftwerk UnioG A n Hammerbacherstraß+ 2 1 e 14 D-8520 Erlangen Tel. 09131-18-0 Telex 62929-0~~

~~Mr. G. GROOS Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. G. HAGSTOTZ Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheil m Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. H.J. HANTKE Hantke-Energietechnik Pöglschlam A 1 2 g 8080 Fürstenfeldbruck~~

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~~Mr. H.-W. HASSEL NUKEM GmbH. Rodenbacher Chaussee 6 Postfach 110080 D-6450 Hana1 1 u Teled xk 418nu 0 478~~

~~527 Mr HAWICKHORS. .W T Kraftwerk UnioG nA Hammerbacherstr. 12 + 14 D-8520 Erlangen Tel. 09131-18-0 Telex 62929-0~~

~~Mr. W. HEIT Rodenbacher Chausse6 e Postfach 110080 D-6450 Hanau 11 Telex 4184 780 nuk d~~

~~Mr. M. HENRICH Vereinigte Elektrizitätswerke Westfalen AG Rheinlanddamm 24 D-4600 Dortmund Tel. (0231) 438-3167 Telefax 0231-438-2147~~

~~Mr. K.-H. HERZBERGER Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße8 D-6800 Mannheiml Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. K. HEUSINGER Kraftwerk Union Hammerbacherstraße 12-14 D-8520 Erlangen Tel. 09131-18-0 Telex 62929-0~~

~~Mr HOFFMAN. .D N Gesellschaft für Reaktorsicherheit (GRS) GmbH Schwertnergasse l D-5000 Köln l TeleD xS 221GR 3 412~~

~~Mr. K. HOFMANN Technischer Ueberwachungs-Verein e.V. Steubenstraße 53 D-4300 Essen l Tel. 0201/8250 Telex 8579680~~

~~Mr. H. HORST Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. H. HORST Tilsiter Str. 35 D-6944 Hensbach~~

~~528 HUEBE. H . LMr Interatom GmbH. Dep. T2 Friedrich-Ebert-Straße D-5060 Bergisch Gladbac1 h Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

~~Mr. X. HUANG at present; Kernforschungsanlage Jiilich GmbH. Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 83355D A 6KF~~

~~Mr. G.P. IVENS Arbeitsgemeinschaft Versuchsreaktor AVR GmbH. Postfach 1411 D-4000 Düsseldorf l Telex 211 4285~~

~~Mr. W. JAHNS Interatom GmbH. Dep. PF12 Friedrich-Ebert-Straße D-5060 Bergisch Gladbacl h Tel. 02204/84-0 Telex 8877 iag) 845 (d l~~

~~Mr. W. JANSING Interatom GmbH. Dep. T31 Friedrich-Ebert-Straße D-5060 Bergisch Gladbacl h Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

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~~KASPE. A . RMr Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~529 Mr. B.H.H. KASSEBOHM Stadtwerke Düsseldorf AG Postfach 1136 D-4000 Düsseldorl f Tele 4411 21 x 8~~

~~Mr KATSCHE. .W R Kernforschungsanlage Jülich GmbH. ISF Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-0 Telex 462041~~

~~Mr. N. KIRCH Kernforschungsanlage Jülich GmbH. Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-554 Telex 83355D A 6KF~~

~~Mr. D. KLEIN Interatom GmbH. Dep. T231 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

~~Mr KLEI. .W N Kraftwerk UnioG nA Department BF 14 Hammerbacherstr4 1 . D-8520 Erlangen Tel. 09131-18-3811 Telex 62929-0~~

~~KLEINE-TEBB. A . Mr E Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheiml Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~KLUSMAN. A . Mr N Ruhrkohle Oel und Gas GmbH. Gleiwitzer Platz3 D-4250 Bottrop Telex 8578 kom949 od~~

~~Mr. W. KOHL Brown, Boveri & Cie.AG Kallstadterstraße l Postfac1 35 h D-6800 Mannheim 31 7 10 1 41 Tele 2 x46~~

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~~Hr. K. KRUEGER Arbeitsgemeinschaft Versuchsreaktor AVR GmbH. Hambacher Forst D-5170 Jülich Telex 833598~~

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~~LANG. G . E Hr Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße8 D-6800 Mannheiml Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Hr. G. LOHNERT INTERATOH GHBH. Dep. PF211 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 8878457 IAG) (D L~~

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~~531 MAUERSBERGE. R . Hr R Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straß8 e D-6800 Mannheil m Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. Kh. MARCH Kraftwerk UnioG A n Dept. V 49 Hammerbacherstr. 14 D-8520 Erlangen Tel. 09131-18-3811 Telex 62929-0~~

~~Mr. M. MARK-MARKOWITSCH Kernforschungsanlage Jülich GmbH. P.O191x Bo . 3 D-5170 Jülich Tel. 06-02461.61-0 Telex 83355D A 6KF~~

~~MARNER. C . Mr T Boare Membeth df o r Arbeitsgemeinschaft Versuchsreaktor Luisenstraße 105 D-4000 Düsseldorf~~

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~~MOORMAN. R . Mr N Kernforschungsanlage Jülich GmbH. ISF Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 83355D A 6KF~~

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~~NICKE. H . LMr Kernforschungsanlage Jülich GmbH. Institute for Reactor Materials 191x Bo P.O3. D-5170 Jülich Tel. 06-02461.61-0 D A KF 6 55 Tele3 83 x~~

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~~MrSCHROEDE. G . R Steag AG Bismarkstr. 54 D-43 Essen~~

~~H.J. Mr . SCHROETER Bergbau-Forschung GmbH Abt. Kohlechemie Grundprozesse Franz-Fischer-Weg 61 Postfach 13 01 40 D-4300 Essen 13 0 BER83 Tele D G 7 x85~~

~~Mr. F. SCHUBERT Kernforschungsanlage Jülich GmbH. Institute for Reactor Materials P.O191x Bo .3 D-5170 Jülich Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~536 Ms. M.L. SCHUELLER Kernforschungsanlage Julien GmbH. Projektträger HTR Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-554 Telex 833556 KFA D~~

~~Mr. SCHU.H G Bundesministerium für Forschung und Technologie Postfach 200706 D-5300 Bonn 2~~

~~Mr. R. SCHULTEN Director IRE Kernforschungsanlage Jülich GmbH. Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-554 Telex 83355D A 6KF~~

~~Mr. H. SCHUSTER Kernforschungsanlage Jülich GmbH. Institute for Reactor Materials P.O191x Bo . 3 D-5170 Jülich Tel. 06-02461.61-554 Telex 833556 KFA D~~

~~Mr. D. SCHWARZ Vereinigte Elektrizitätswerke Westfalen AG Theinlanddam4 2 m D-4600 Dortmund~~

~~Mr. W. SEELE Brown, Boveri & Cie AG Dept. Switchgear Postfac1 35 h Kallstadter Straße l D-6800 Mannheim Telex 462 411 202~~

~~Mr. J. SINGH Kernforschungsanlage Jülich GmbH. IRE Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 83355D A 6KF~~

~~Mr. B. STAUCH Kernforschungsanlage Jülich GmbH. IRE Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 83355D A 6KF~~

~~5-37 STEHL. H . EMr Interatom GmbH Dep. T311 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

~~Mr STEINWAR. .W Z Interatom GmbH Dep. PF113 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 887 7 iag) 845 (d l~~

~~Mr. M. STELZER Ruhrkohle Oel und Gas GmbH. Gleiwitzer Platz3 D-4250 Bottrop~~

~~Mr STOEL2. .D L Brown, Boveri & Cie AG Mannheim, Germany Nuclear Energy Dept. Postfach 351 Kallstadter Straße l D-6800 Mannheim 31 Telex 462 411 107~~

~~Mr. H.W. STRAUß Rheinische Braunkohlenwerke AG Stüttgenweg 2 D-5000 Köln 41 Telex 888301d w rb 1a~~

~~Mr. H. TEUBNER Interatom GmbH. Dep. T311 Friedrich-Ebert-Straße D-5060 Bergisch Gladbach l Tel. 02204/84-0 Telex 887 7 iag) 845 (d l~~

~~Mr. E. TEUCHERT Kernforschungsanlage Jülich GmbH. IRE Postfach 1913 D-5170 Jülich l Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~Mr THEYMAN. .W N Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße8 D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~538 Mr VALSA. .M N Kernforschungsanlage Julien GmbH. IRW Postfach 1913 D-5170 Jülicl h Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~VERPONDER. K . Hr N Kernforschungsanlage Jülich GmbH. P.O. Box 1913 D-5170 Jülich Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~Mr. W. WACHHOLZ Hochtemperatur-Reaktorbau GmbH. Gottlieb-Daimler-Straße 8 D-6800 Mannheim l Tel. 0621/4511 Telex 462041; Telefax 0621451599~~

~~Mr. J. WALDMANN Kraftwerk UnioG A n Departmen4 7 tR Hammerbacherstraße 14 D-8520 Erlangen Tel. 09131-18-0 Telex 62929-0~~

~~WEISBROD. I . Mr T Interatom GmbH. Dept. PF1 Friedrich-Ebert-Straße D-5060 Bergisch Gladbacl h Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

~~Mr WIL. .M L Interatom GmbH. Dep. PF114 Friedrich-Ebert-Straße D-5060 Bergisch Gladbacl h Tel. 02204/84-0 Telex 8878 457 iagl (d)~~

~~MrWOL. L . F Kernforschungsanlage Jülich GmbH. IRE 191x Bo P.O3. D-5170 Jülich Tel. 06-02461.61-0 Telex 833556 KFA D~~

~~WOHLE. J . RMr Hochtemperatur-Kernkraftwerk GmbH. Siebenbeckstr0 1 . D-4701 Hamm-Uentrop~~

~~539 Hr. E. ZIERMANN Arbeitsgemeinschaft Versuchsreaktor AVR GmbH. Hambacher Forst D-5170 Jülich Telex 833598~~

~~Mr. I.L. ZIERMANN AV Gmb- R H o Kernforschungsanlagc/ e Jülich Postfach 1913 D-51700 Jülich Telex 83355D A KF 6~~

~~INDIA~~

~~Ms. K. BALAKRISHNAN Bhabha Atomic Research Centre Reactor Engineering Division Trombay Bombay-400 085 Tel. 523321 Telex 281 Codb A 7 e Bare~~

~~INDONESIA~~

~~Mr DJOKOLELON. .M O National Atomic Energy Agency . K.HJl . Abdul Rochim Kuningan Barat Mampang Prapatan Jakiarta Selatan Telex 46354 BATAN JKT~~

~~Mr. I.R. SUBKI National Atomic Energy Agency . K.HJl . Abdul Rochim Kuningan Barat Mampang Prapatan Jakiarta Selatan Telex 46354 BATAT NJK~~

~~IRAQ~~

~~D.K. Mr . AL-DABAGH Atomic Energy Commission Tuwaitha-Baghdad 5 76 P.Ox Bo . Baghdad, Iraq~~

~~ISRAEL~~

~~Mr. J. ALTER Israel Atomic Energy Commission P.O. Box 7061 Tel-Aviv 61070 TeleL xI 3345F P T 0S~~

~~540 Mr. W.S. DUKHAN NRCN-IAEC 900x Bo P.O1. Beer-Sheva 84190 Telex 5363 RTKML I G~~

~~GREENSPA. E . Mr N Israel Atomic Energy Commission P.O. Box 7061 Tel-Aviv 61070 Telex 33450 ST PF IL~~

~~Mr. A. SEROUSSI Israel Atomic Energy Commission P.O. Box 7061 Tel Aviv 61070 Telex 33450 ST PF IL~~

~~ITALY~~

~~GARRIB. S . Mr A Politechnical University Milano~~

~~Mr. 0. MODONESI ENEA, CRE Cassaccia, Postal Box 2400 1-00100 Rome Tel. 040 6 6948 Telex 613296~~

~~JAPAN~~

~~Mr. Ryuichi ARAKI Chiyoda Chemical Engineering & Construction Co., Ltd. 1-12 Tsurumichuo 2-chome Tsurumi-ku Yokoham0 a23 Telex 47726 chiyo J~~

. OsamMr u BABA Japan Atomic Energy Research Institute (JAERI) Tokai Research Establishment Dept Powef .o r Reactor Projects Tokai-mura, Naka-gun Ibaraki-ken, 319-11 Tel. 02928-2-5764 Telex 3632339 JAERJ IF

Toshikaz. Mr u HAYASHI Japan Atomic Energy Research Institute (JAERI) Tokai Research Establishment Dept. of Power Reactor Projects Tokai-mura, Naka-gun Ibaraki-ken, 319-11 Tel. 02928-2-5764 Telex 3632339 JAERJ IF

~~541 MEXICO~~

~~J.B. Mr . MORALES Institu Investigacionee td s Eléctricas Apartado Postal 475 Centro-Cuernavaca, Mor. 62000 Mexico, D.F. Tel. 514-8004 514-8259 Telex 17776352 HEMME~~

~~THE NETHERLANDS~~

~~Mr. J.M. van den BRINK NUCON Engineering & Contracting B.V. P.O 402x .Bo 6 100 AmsterdaA A 9 m Telex 11996~~

~~. J.B.MMr HAAe .d S Netherlands Energy Research Foundation Westerduinweg 3, P.O. Box l 175 G PetteZ 5 n Telex 57211 REACL PN~~

~~Mr. G.G. van UITERT Ministry of Economic Affairs 2010P.Ox Bo .1 Hague 250Th eC 0E Telex 31099 B ECZA NL~~

~~POLAND~~

~~Mr. Edward OBRYK Institute of Nuclear Physics Ul. Radzikowskiego 152 31-342 Krakow Tel. 3370222 ext. 280 Telex 032246L P j 1if~~

~~PORTUGAL~~

~~Ms. I.M. CANHAO RORIZ Gabinete de Protecçao e Segurança Nuclear Ministerio da Industria Av. Republica 45 - 6° 1000 Lisbon Telex 14344 NucseP g~~

~~Mr. F. CARDEIRA Laboratori e Physicod n Engenhariae a Estrada Nationale 10 Sacavam Telex 12727 NUCLABP~~

~~542 SPAIN~~

~~Consuel. Ms o PERE MORAL SDE L Représentante Técnico c/o Consej Seguridae od d Nuclear (CSN) International Affairs Sor Angela de la Cruz, 3 28020 Madrid, Spain Tel. 4561812 Telex 45869 CSNE M~~

~~FERNANDE. C . Mr Z PALOMERO Central Nuclear de Vandellos Directo Vandelloe d r I s Hospitalet del Infante Tarragona Tel. 823050 Telex 56430~~

~~ROMANIA~~

~~Mr. I. NEAMU Institute of Physics and Nuclear Engineering B.P. 35 Bucarest Tel. 00400-807040 Telex 11397 CSEN R~~

~~SWEDEN~~

~~Mr. C.U. RUNFORS Studsvik Energiteknik AB S-611 82 Nyköping Tel. 0155/210 00 Telex 64013 STUDS Telefax 0155-63044~~

~~SWITZERLAND~~

~~Mr. P.O. BURGSMUELLER Gebrüder Sulzer AG Abt. 0350 CH-8401 Winterthur 0 3 0 06 Tele 6 89 x~~

~~HOL. H M. Mr General Atomic Europe Dorfstraße 4 CH-8126 Zumikon-Zurich Tel. 01-9180505 Telex 53499~~

543 Mr. G. SARLOS Schweizerische Interessen- gemeinschaf Wahrnehmunr zu t g gemeinsamer Interessen an der Entwicklung nuklearer Technologien (IGNT) CH-5303 Würenlingen Telex H C 5371 R 4EI Tel. 056/992742

~~USSR~~

V.N. Mr . GREBENNIK I.V. Kurchatov Institute of Atomic Energy Ulits2 4 a Kurchatova 340x P.OBo 2. 123182 Moscow Tel. (095) 1965158 (State Committee on Utilization of Atomic Energy Tel. 2331718, Telex 411888 MEZON SU)

~~Mr. I.V. CHROMOV USSR State Atomic Energy Committee Staromonetny6 2 , Moscow 109180 USSR~~

~~Mr. I.M. IONOV State Committee on the Utilization of Atomic Energy Staromonetny peureulok6 ,2 Moscow Zh-180 Tel. 2331718 Telex 411888~~

~~L.N. Mr . PERMERKOV State Committee th n o e Utilization of Atomic Energy Staromonetny peureulok, 26 Moscow Zh-180 Tel. 2331718 Telex 411888~~

~~UNITED KINGDOM~~

~~. J.RMr . ASKEW UK Atomic Energy Agency E511, Risley Warrington, CheshirT 6A 3 eWA Tel. 092531244 Telex 629301 atomry g~~

~~Mr. B. KEEN National Nuclear Corporation Limited Booths Hall, Chelford Road Kuntsford, Cheshire WA1Z 68Q Telex 666000~~

~~544 PEXTO. A . NMr SSEB Cathcart House Spean Street GlasgowE 4B 4 ,G4 Tel. 041-637-7177 Telex 777703~~

~~Mr. P.T. SAWBRIDGE Central Electricity Generating Board Berkeley Nuclear Laboratories Berkeley, Gloucestershire, GL1B 39P Telex 43227~~

~~UNITED STATES OF AMERICA~~

~~BRE. L Y. Mr Manager of Nuclear Engineering Public Servic f Colorado e o 2420 West 26th Avenue Denver, CO 80211~~

~~Mr. J.L. HELM President Proto-Power Cooperation Groton, Connecticut 06340 Telefax 203/4468292~~

~~Mr. J. JONES Oak Ridge National Laboratory X P.Ox Bo . Oak Ridge, Tennessee 37831 Tel. (615) 574-0377, Telefax 624-0377~~

~~Mr. A. MILLUNZI Division of HTGR Offic Nucleaf o e r Energy DOE Washington Mail Stop-B-107/GTN Washington, D.C. 20545 Tel. 900-1-301353-2947 Telex 710-828 0475 Telefax 9001-301-353-3870/3888~~

~~Mr. A.J. NEYLAN GA-Technologies Inc. P.O. Box 85608 San Diego A 9213,C 8 Tel. (6195 )45 Telex 695065~~

~~YUGOSLAVIA~~

~~MrVARAZDINE. .Z C Institut za Elektroprivredu Proleterskih brigad7 3 a 41000 Zagreb Tel. 041/530-604 or 041/513-822/338 Telex 2215 U INSTE4Y P~~

~~545 INTERNATIONAL ORGANIZATIONS~~

~~Commissio Europeae th f o n n Communities (CEO~~

~~Mr. C. ViVANTE 200, rue de la Loi B-1049 Bruxelles Tel. 235 11 11 Telex COMEU B 21877~~

~~Organizatio Economir nfo c Co-operatio Developmend nan t (OECD)~~

~~Mr. S. HORIUCHI Deputy Director for Nuclear Science and Development OECD Nuclear Energy Agency , boulevar38 d Suchet F-75016 Paris 0 5 6 9 4 Tel52 . AEN/NE8 66 Tele 0 A63 x~~

~~International Atomic Energy Agency (IAEA)~~

~~KUPIT. J . ZMr Divisio Nucleaf o n r Power Wagramerstraft5 e 0 10 P.Ox Bo . A-1400 Vienna Tel. 2360-ext.2814 through-dialing Telex 1-12645 Telefax 222-230184~~

~~INTERNATIONAL PROGRAMME COMMITTEE~~

~~AUSTRIA~~

~~Mr. G.F. REITSAMER OFZ SSeibersdor- f~~

~~FEDERAL REPUBLI GERMANF CO Y~~

~~Jülic- H PT h BALTHESE. E . Mr N~~

~~MrPOH. .H L BMF BonT- n~~

~~Jülic- A KF h . SCHULTER . Mr N~~

~~546 UNITED STATE AMERICF SO A~~

~~Mr. A. MILLUNZI DOE - Washington~~

~~COMMISSION OF THE EUROPEAN COMMUNITIES~~

~~MrVIVANT. .C E~~

~~INTERNATIONAL ATOMIC ENERGY AGENCY~~

~~Mr. J. KUPITZ~~

~~547~~