Design and Development of Gas Cooled Reactors with Closed Cycle Turbinesgas

IAEA-TECDOC-899

Design and development of gas cooled reactors with closed cycle turbinesgas

Proceedings of a Technical Committee meeting held in Bei/ing, China, 30 October - 2 November 1995

INTERNATIONAL ATOMIC ENERGY AGENCA / Y The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies of these reports can be obtained from

IN IS Clearinghouse International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria

Orders shoul accompaniee db prepaymeny db f Austriao t n Schillings 100,- in the form of a cheque or in the form of IAEA microfiche service coupons which may be ordered separately from the IN IS Clearinghouse. originatine Th g Sectio f thino s publicatio IAEe th An i was- Nuclear Power Technology Development Section International Atomic Energy Agency Wagramerstrasse 5 0 10 x P.OBo . A-1400 Vienna, Austria

DESIG DEVELOPMEND NAN COOLES GA F TDO REACTORS WITH CLOSED CYCL TURBINES EGA S IAEA, VIENNA, 1996 IAEA-TECDOC-899 ISSN 1011-4289 IAEA© , 1996 Printed by the IAEA in Austria August 1996 FOREWORD

lons Iha tg been recognized that substantial generatioe gainth n si f electricitno y from nuclear e obtainefissiob n ca nd throug e directh h t s turbincouplin ga a hig a o t ehf o g temperature helium cooled reactor. This advanced nuclear power plant is unique in its use of the Brayton cycle to achieve a net electrical efficiency approaching 50% combined with the attendant features of low initial capital costs due to plant simplification, public acceptance from the safety attributes of the high temperature gas cooled reactor (HTGR), and reduced radioactive wastes.

e TechnicaTh l Committee Meeting (TCM d Worksho an )e Desig th d n o an pn Developmen Cooles Ga f o dt Reactors with Closed Cycl Turbines eGa convenes swa d within the frame of the International Working Group on Gas Cooled Reactors as part of the IAEA nuclear power technology development programme.

Technological advances over the past fifteen years in the design of turbomachinery, recuperator d magnetian s c bearings provid e potentiath e quantua r fo l m improvemenn i t nuclear power generation economic HTGe th f so R throug e witus closeha e hth d cycls ega turbine. Enhanced international co-operation among nationa cooles ga l d reactor programmes in these common technology areas could facilitate the development of this nuclear power concept thereby achieving safety, environmental and economic benefits with overall reduced development costs Workshod an . ThiM convenes psTC wa provido dt opportunite eth o yt review and examine the status of design activities and technology development in national HTGR programmes with specific emphasi closee th n dso cycl turbines identifo et ga d an ,y pathways which take advantage of the opportunity for international co-operation in the developmen f thio t s concept. EDITORIAL NOTE

preparingIn this publication press,for IAEAthe staffof have pages madethe up fromthe original manuscripts submittedas authors.the viewsby The expressed necessarilynot do reflect those governmentsofthe nominating ofthe Member nominating the States of or organizations. Throughout textthe names Memberof States retainedare theyas were when textthe was compiled. Theof use particular designations countriesof territoriesor does imply judgementnot any by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the pan of the IAEA. The authors responsibleare havingfor obtained necessarythe permission IAEAthe to for reproduce, translate or use material from sources already protected by copyrights. CONTENTS

7 SUMMARY

SUMMARY OF NATIONAL AND INTERNATIONAL ACTIVITIES IN GAS COOLED REACTORS (Session 1)

Construction of the HTTR and its testing program for advanced 5 1 HTGR development T. Tanaka, O. Baba, S. Shiozawa, M. Okubo, K. Kumtomi Progress of the HTR-10 project 25 D. Zhong, Y. Xu State of HTGR development in Russia 31 V.F Golovko, Kiryushtn,I. A N.G. Kodochigov, N.G. Kuzavkov High temperature reactor development in the Netherlands 47 Heekvan .1. A French activities on gas cooled reactors 51 D. Bastten Status of GT-MHR with emphasis on the power conversion system 55 A.J. Neylan, F.A. Silady, B.P. Kohler, Lomba,D. RoseR. HTR plus modern turbine technology for higher efficiencies 67 H. Bamert, K. Kugeler

DESIG HTGRF NO s WITH CLOSED CYCL TURBINES EGA S (Sessio) n2

Design of indirect gas turbine cycle for a modular high temperature 5 8 s coolega d reactor Z. Zhang, Z. Jiang Conceptua5 9 l desig heliuf no turbins mga r MHTGR-Gefo T Matsuo,E. Tsutsumi,M. Ogata,K. NomuraS. Investigation of GT-ST combined cycle in HTR-10 reactor 111 Z. Gao, Z. Zhang, D. Wang HTG turbine-moduls Rga 1 12 e demonstration tesproposaa - t l E Takada, Ohashi,K. Hayakawa,H. Kawata,O. KobayashiO.

LICENSING, FUEL AND FISSION PRODUCT BEHAVIOUR (Session 3)

A study of silver behavior in gas-turbine high temperature gas-cooled reactor 131 K. Sawa, S. Shiozawa, K. Kumtomi, T. Tanaka COMEDIE BD1 experiment Fission product behaviour during depressunzation transients 145 R. Gillet, D. Brenet, D.L. Hanson, O.F. Kimball 7 15 Licensing experienc HTR-1e th f eo 0 test reactor Y. Sun, Y. Xu

GAS-TURBINE POWER CONVERSION SYSTEM DEVELOPMENT (Session 4)

Development of compact heat exchanger with diffusion welding 165 K Kumtomi, T. Takeda, T. Hone, K. Iwata Summary repor technican o t l experiences from high-temperature helium 7 17 turbomachinery testin Germann gi y WeisbrodtLA

WORKSHOP (Sessio) n5

Pebble bed modular reactor - South Africa 251 M. Fox, E. Mulder IAEe Rolth f gas-cooleAn ei o d reacto7 r developmen25 applicatiod an t n J. Cleveland, L. Brey, J. Kupitz

List of Participants 271 SUMMARY

The Technical Committee Meeting (TCM) and Workshop on Design and Development s CooleoGa f d Reactors with Closed Cycl s TurbineGa e s hel Beijingn wa i sd , People's Republic of China, from 30 October to 2 November 1995. The meeting was convened by the recommendatioe IAEth n Ao Internationas it f no l Working Cooles GrouGa n dpo Reactors (IWGGCR Instituthostes e th wa y dd b Nucleaf e)o an r Energy Technology (INET), Tsinghua University. The meeting was organized to review the status and technology development activities for high temperature gas cooled reactors with closed cycle gas turbines (HTGR-GT), an identifo dt y pathway opportunitf so r internationayfo l co-operatio developmene th n i f o t common systems and components. It was attended by twenty-three participants and observers from eight countries (China, France, Germany, Japan, the Netherlands, the Russian Federation, South Africa and the United States of America). Sixteen papers were presented by the participants on behalf of their respective countries. Professor Z. Wu, Director of INET, chaired the meeting. Each presentation was followed by general discussion within the area covere papere th y db .

INEToure th f To s research facilitieHige th hd Temperaturan s e Reactor (HTR-10) construction site followed the meeting and workshop.

Recent advance turbomachineryn si , magnetic bearing head san t exchanger technology provid potentiae eth r significanfo l t improvemen nuclean i t r power generation economics through use of the HTGR with closed cycle gas turbines. This TCM was used as a forum to share the recent advances in this technology and to explore those technical areas where international co-operation woul e beneficiadb e HTGR-GTth o t l . Area f investigatioo s n included materials development, component fabrication, qualification of the coated fuel particles and fission product behaviour hi the power conversion system, national HTGR development and testing programmes including gas turbine related systems and components.

Summarie f nationaso internationad an l cooles ga l d reactor development programmes included presentations from representatives of Japan, China, the Russian Federation, the Netherlands, France, the United States of America and Germany.

The Japanese High Temperature Engineering Test Reactor (HTTR) is currently under construction at the Japan Atomic Energy Research Institute (JAERI), Oarai Research Establishment site. This is a 30 MWth reactor with a maximum outlet temperature of 950°C. Construction of the HTTR began in March, 1991, with initial reactor criticality scheduled for lat reactoe e 1997Th . r buildin containmend gan t vessel, includin maie gth n components such as the reactor vessel, intermediate heat exchanger, hot gas piping and the reactor support structure are now installed. The HTTR project is intended to establish and upgrade the technology basis necessar advancer yfo d HTGR development heae Th .t utilization systems planned for demonstration with this reactor are currently under consideration as part of the IAEA Co-ordinated Research Programm Desige th Evaluatiod n eo nan f Heano t Utilization Systems for the HTTR. The gas turbine and CO2 and steam reforming of methane are considered as first priority heat utilization system candidates. Safety demonstration tests on the HTTR are also planned to confirm the inherent safety features of the HTGR.

Construction continue HTR-1e th INE e n th so t T0a site outsid f Beijingeo , China. This pebble-bed helium cooled test reacto beins i r g supporte governmene th y db f Chinao t e Th . construction permit for this plant was issued by the national safety authority at the end of 1994 and construction bega Junen i , 1995. First criticalit schedules yi r 1999dfo . maie Mosth nf o t components, including the spherical fuel elements, are to be manufactured in China. Development of the HTGR in China is being undertaken as both a source of electrical generatio supplemena s na theio t t r water cooled universaa reactor s a d san l heat sourco et provide process heat for various industrial applications. There will be two phases of high temperature heat utilization fro HTR-10e mth firse Th t. phase will requir ereactoa r outlet temperature of 700°C. with a conventional steam cycle turbine in the secondary loop. The second phase will utilize a core outlet temperature of 950°C. primarily to investigate a steam cycle/gas turbine combined cycle system where the gas turbine and the steam cycle are independently parallel in the secondary side of the plant.

HTGa d Russiha R s developmenaha t programme spanning ove years0 3 r . Detailed design VGR-50e th f so hav, M VG-40e beeVG d n0an completed feasibilitA . 5 y21 studa r yfo MWth (termed VGMP) plan performes wa t 199n di 1992d 1an . Presently, development work withi frame nth f internationaeo l co-operation with Genera takins i A lg Atomic US plac e th e f so turbins ga e oneth modular helium reactor (GT-MHR). This plant would hav thermaea l power rating of 550-600 MWth with a net electrical efficiency of approximately 46%. The Russian gas cooled reactor strategy include pebblsa modulad ebe r HTGR wit thermaha l power level of about 200MW aime procest da s heat production prismatia d an , c core modular HTGR with turbins ga r electricaa efo l generation relativele th n I . y long histor f HTGyo R development, Russi carries aha significant dou t development wor HTGn ko R components includin hige gth h temperature heat exchanger, steam generator, circulator, fuel, graphit d structuraan e l materials.

Research and development activities on the HTGR began in the Netherlands in 1993. Their activities now concentrate on the development of a small pebble bed HTGR for combined heapowed tan r production wit hclosea d cycl turbines ega concepe Th . bases i t n do peu-a-pee th u design from Forschungszentrum Julich GmbH (KFA), wit overaln ha l design philosophy of achieving significant plant simplicity. An independent neutronics and thermal hydraulics computer code syste s mbeini g developed with international co-operation i n performing benchmark calculations e objective Th Netherlandse . th f o s ' HTGR activities include restoring social suppor their fo t r nuclear programme, achieving commercial viability of the HTGR and to make the market introduction of the HTGR economically feasible with limited development and first-of-a-kind costs.

France has built and operated eight natural uranium graphite moderated CO2 cooled reactor exported f servic f theso Spainsan o t o t theil d e eou e an Al don .reactor r w no e sar presen cooles ga t d reactor related activities focu studn so developmend yan techniquef o t f so plant decommissioning.

cooles ga e d reactoTh havA r activitieeUS focuse e conceptuae th th n sn i do l desigf no the GT-MHR. Significant progress has been made on this concept since its introduction in 1993. However, as part of a broad based decrease in government funding, the United States Department of Energy (DOE) terminated this programme in July, 1995. The GT-MHR programme work performe dato dt e include evaluatioe sth technicae th f no l issues centering e poweoth n r conversion component se turb th suc os a h machine, plate-fin recuperator, magnetic bearings and helium seals. A key event in die United States GT-MHR programme was the review by Electric Power Research Institute utility representatives in June, 1995, which focuse powee th n dro conversion system reviee Th . w team concluded partn i , , thae th t GT-MH bola innovativ d s Ri dan f emergino e eus g technologies that hav potentiae eth r fo l

8 much improved efficiency, simplification and reduction of costs and wastes. This review team also acknowledged thacurrene th t t United States GT-MHR conceptual design needs substantial detailed design, experimental verification and licensing effort before it can be viewed with confidenc breakthrouga s ea h technology that would provid technicallea economicalld yan y viable future optio potentiar nfo l utility customers utilite Th . y representatives foun showo dn - stoppers that would prevent the GT-MHR from eventually meeting regulatory safety goals and achieving the technical and economic goals set by the DOE and the designers.

In Germany s coolega e dth , reactor programm t KFAea , Jiilich s evaluatini , e gth technica s turbinl ga lin f eo e combined cycle units aime t achievinda g higher efficiencf o y electrical production. A comparative study of efficiency potential has shown that an electrical generation efficiency of 54.4% is achievable with a gas turbine inlet temperature of 1050°C. e mechanisTh f mfissioo n product release from spherical HTGR fuel element s beeha s n reviewe evaluato dt technicae eth l feasibilit f increasinyo coolane gth t outlet temperaturf eo pebble bed reactors. Good operating experiences at the AYR plant with a mean outlet temperatur f 950°Ceo , together with other experimental result proposald san increaso st e eth retention capability of the HTGR fuel elements, is regarded by German experts as reasons for possible realizatio reactof no r outlet temperatures approaching 1050°C.

Both China and Japan are investigating conceptual designs featuring the use of gas turbines. In China, the focus is on a 200 MWth pebble bed MHTGR with an indirect gas turbine cycle utilizing heliu primare th ms a y coolan nitroged coolantan e turbinee th th s n a o t . This plant would have a core outlet temperature of 900 °C and is expected to have an electrical generation efficienc abouf yo t 48% Japanese .Th investigatine ear direce gth t cycle helius mga turbin r botfo e hpracticaa experimentan MWt0 a l 45 d uni f han o t l uni f 120o t 0 MWth. Comparisons have been single madth f eo shaft heliu turbins mga e type employin axialn ga - flow compresso twia d n ran shaf t type employin centrifugae gth l compressor single Th . e shaft advantagee th unis ha t bettef so r structur controd ean l designtwie th nt shafbu , t unihighes ha t r efficiency. Investigation now centers on the safety features and startup characteristics of each design.

Testing of the power conversion components as an integral system utilizing a non- nuclear heat source is seen as a prerequisite prior to connection to the reactor. The therrnodynamic performance of the power conversion system can be demonstrated and verified fulo t l temperatur speed ean d condition t considerablsa y lower pressures the plane nth t design pressure proposaA . l from Japan utilize electrican sa l heateprimare th s ra y energy supplye Du . to the high efficiency of the power conversion system, an outside power supply of only approximately 10% of the thermal plant rating would be required for a test of this nature.

Fission product behaviour for the direct cycle gas turbine HTGR is a very important aspec considerinn i t desige g th maintenanc d nan thif eo s plant turbine Th . e firse bladeth te sar component encountere t primarho e th y y coolandb t helium upon leavin reactore gth . Silver, wit tendencs hit higt plata o y t t h eou temperatures iodine-13d an , caesium-13d 1an 7 with plate ou t lowea t r temperatures e necessarar , y design consideration orden i s minimizo t r e th e radiological aspect thin so s behaviou e plantth s wels a A . t silvef l ro no know s i re th thas n a f o t noble gases, iodine and cesium, this has been an area under investigation by Japan in their evaluatio GT-MHRe th f no . Also irradiation a , n test program carrieFrancy b t theidn ou ei r COMEDI loo1 pEBD included securing dat fission ao n product release fro mfuea l element, fission o d n plat an n produco t eou t liftoff. After steady state irradiation looe s subjecteth , pwa d t oseriea in-sitf so u blowdown sheat sa r ratios ranging betwee5.6d an . 7 Evaluation0. e th f no results with regard to fission product profiles show that the plate out profiles depend on the fission product chemistr thad depressurizatioe yan th t n phasesignificano t d sle t desorptiof no iodine-131, but had almost no effect on other fission products such as silver-110m, caesium- 134 and -137, and tellurium-132.

The HTR-10 is the first high temperature gas cooled reactor to be licensed and constructed in China. The purpose of this reactor project is to test and demonstrate the technolog safetd yan y feature advancee th f so d modular high temperature design licensine Th . g process for this first HTGR in China represented challenges to both the regulator and the designer e maximuTh . m fuel temperature unde accidene th r t conditio f completno e losf so coolant was limited to values considerably lower than the safety limit set for the fuel element. Conversely reactoe th , r incorporates many advanced design feature aree passivth f ao n i s d ean inherent safety, and it is presently a worldwide issue as to how to properly treat these safety features in the licensing review process. Some of the main safety issues addressed in the licensing procedure included fuel element behaviour, source term, classification of systems and components containmend an , t design HTR-1e Th . licenses 0wa tesa s dta reactor rather than a power reactor licensine Th . g experience Chinn si thir afo s reactor coulf significano e db t reference valu futurn ei e HTGR licensing efforts worldwide.

closee Th d cycl turbins ega e design generally incorporate platsa typn efi e recuperator in the power conversion system. The normal means of connecting the plates and fins is by brazing which may not have long term reliability. Diffusion welding of the plates to the fins is currently being developed for these recuperators. Tensile and creep strength in the diffusion welds, especially in high temperature applications, has shown to be superior to brazing. Early testindiffusioe th f go n welds als indicates oha d high reliability.

Because of the strong relevance of the IAEA's HTGR programme to current developmental activities associated with the gas turbine, the IAEA Nuclear Power Technology Development Section initiate tasda assimilato kt desige e th operationa d nan l experiencef so high temperature helium driven turbomachinery testing previously carried out in Germany. The information assembled as the result of this effort was reviewed by the author at the TCM and is included herein.

A comprehensive programm initiates ewa 196n di Germann 8i researce th r d yfo han development of a Brayton (closed) cycle power conversion system. The programme was to ultimatel a HTG e r electricaus yRfo l generation with heliue workinth s ma g fluid. This programme continued until 1982 and involved two experimental facilities. The first was an experimental co-generation power plant (district heating and electrical generation) constructed and operate utilitye th y db , Energieversorgung Oberhausen (EVO). This consiste fossia f do l fired heater, helium turbines, compressor related san d equipment secone Th . de facilitth s ywa High Temperature Helium Test Plant (HHV) for developing helium turbomachinery and component t KFAsa derives , wa Jiilich V dheae froHH Th t. electrin sourcme a th r efo c motor driven helium compressor broaA . d experimental development base concerning helium turbines, compressors, hot-gas duct, high temperature materials, recuperators, fuel elements, graphite, etc., was performed within the frame of this programme in order to assure the feasibilit f thio y s technology. Positiv d negativan e e experiences were gained from both facilities e dynamicaTh . l turbomachinperformancV HH e th s patentl f o ewa e y excellent, whereas the EVO machine at first showed insufficient dynamical behaviour. This behaviour was significantly improved with shaft and bearing modifications. However, the power deficit turbomachine ofth e overcome coulb t dno e without significant rebuildin r exchanggo e th f eo

10 machine. Modification correcteV HH l ingrese doi th excessivd o st san e leakage problemse Th . positive experiences achieved with both facilities after overcoming initial deficiencies included excellent performance of the gas and oil seals, hot-gas ducting, turbomachine cooling, operation of the helium purification system and of the instrumentation and regulation systems.

workshoe Th p followinfocuseM TC identifyinn de o gth g pathway r internationasfo l co-operation in the development of common areas of the HTGR-GT including plant safety, power conversion system and component design and fuel and fission product behaviour. Included was a summary of the current technological and economic investigation by Eskom, the electrical utilit Soutf yo h Africa, int possible oth e deploymen f smalo t l («100 MWe) closed cycle helium drive s turbinega n o augment s t their electrical system capability. Under investigatio utilizatioe th pebbls ne i th modulad f no e be r reactor couple threa o dt e shaft closed cycle gas turbine power conversion system. The initial technical/economic evaluation is to be f 1996 o decisio A d complete. en e nth wily db l the made nb f whetheeo proceeo t r d inte oth detailed design phase and the subsequent ordering of long lead-time components and fuel. Also addresse e internationae IAE th th a revierole f n As th i o e f wwa do s coolega l d reactor programme. The discussions centered around the general need to co-ordinate this programme worldwide with future emphasi HTGR-Ge th n si T area f defininso g minimum safety related requirement desigd san n basis accidents, developin co-ordinatew ne ga d research programme on power conversion system components, documenting former HTR programmes from countries such as Germany, Switzerland, USA, etc., and establishing a user/utility/vendor association for the promotion of the high temperature reactor.

11 SUMMARY OF NATIONAL AND INTERNATIONAL ACTIVITIES IN GAS COOLED REACTORS

Session1 CONSTRUCTION OF THE HTTR AND ITS TESTING PROGRA ADVANCER MFO D HTGR DEVELOPMENT XA9642774

T. TANAKA, O. BABA, S. SHIOZAWA, M. OKUBO, K. KUNITOMI Departmen f HTTo t R Project, Japan Atomic Energy Research Institute, Ibaraki-ken, Japan

Abstract

Concerning about global warming due to emission of greenhouse effect gas like C02,it is essentially important to make efforts to obtain more reliable and stable energy suppl extendey b ynucleaf o e us dr energy including high temperature heat from nuclear reactors, because it can supply a large amount of energy and its plants emit only little amount of C02 during their lifetime. Hence, efforts are to be continuously devoted to establish and upgrade technologies of High Temperature Gas-cooled Reactor (HTGR) whic n supplhca y high-temperature heat with high thermal efficienc wels ya s a l high heat-utilizing efficiency. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then constructioe th , Higf no h Temperature engineering Test Reactor (HTTR), whics i h an HTGR with a maximum helium coolant temperature of 950°C at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI).

e constructioTh HTTe th R f starteno Marcn i d h 1991, with first criticalitn i y 1997 to be followed after commissioning testing. At present the HTTR reactor building and its containment vessel have been constructed and its main components, such as a reactor pressure vessel n intermediat,a e heat exchanger s pipingga t cord ,ho san e support structures, have been installed in the containment vessel. Fuel fabrication has been started as well.

The project is intended to establish and upgrade the technology basis necessary r advancefo d HTGR developments. Some heat utilization syste plannes i me b o t d connecte HTT e d demonstrateth Ran o t d formee th t r a d secone stagth f deo coret .A present, steam-reforming of methane is the first candidate. Also the HTGR with Gas- Turbine has been studied for assessment of technical viability.

Beside demonstratioe th s heae th t f utilizationo n system JAERe th , I plano t s carr t safetyou y demonstration test confiro t s saliene th m t inherent safety features of the HTGR. In addition material and fuel irradiation tests as upgrading HTGR technologies after attaining rated power will be conducted. Preliminary tests on selected research subjects suc s composita h e materiaC coateZr d dan l fuel developments, have been carried out at high temperature and under irradiation.

15 1. Introduction

A High Temperature Gas-cooled Reactor (HTGR) can supply high temperature heat as hig s 1000°Cha , potentiaa whic s ha h f obtainino l g high thermal efficienc s wela ys a l high heat-utilizing efficiency. It also has excellent features such as high inherent safety, easy operation and high fuel burnup. From the view point of the global environmental protection and diversification of energy usage, non-electrical application of nuclear energy, hydrogen production for example, is very important. Therefore orden i , o establist r upgradd han e technologicath e l basi r HTGRfo d s an s a too f basis o a l e c us alsresearche o t o r higfo s h temperatur neutrod an e n irradiation e Japath , n Atomic Energy Research Institute (JAERI s beeha )n constructing a 30MWt High Temperature engineering Test Reactor (HTTR) at the Oarai Research Establishment firse Th .t criticalit s schedulei y 1997n i d . This report describes present e statuHTTth Rf o sconstructio s testinit d nan g progra r advancemfo d HTGR developments.

2. Present statu f HTTo s R construction

The HTTR plan s composei t a reacto f do r building a spen, t fuel storage building, a machinery building and so on. The reactor building is 48m x 50m in size with two floors on the ground and three under ground. Major components such as the reactor pressure vessel, primary cooling system components etc. are installed in the containment vessel coolinr Ai . g cooline towerth r fo gs syste e locateme rooar th fn do e reactooth f r building constructioe Th . f reactono r building starte Marcn i d h 1991 and has been almost completed. The external view of the reactor building is shown in Photo 1. The major specification of the HTTR are shown in Table 1.

Phot . 1 oExterna e lreacto th vie f wo r building

16 TABL . MAJOE1 R SPECIFICATIO HTTF NO R

Thermal power W M 0 3 Outlet coolant temperature 850'C/950°C Inlet coolant temperature 395°C Fuel Low enriched U02 Fuel element type Prismatic block* Direction of coolant flow Downward-flow Pressure vessel Steel Numbe f maio r n cooling loop 1 Heat removal (paralleC PW d 1Hl an Xloaded ) Primary coolant pressure 4 MPa Containment type Steel containment Plant lifetime 20 years

e blocTh k type fue adoptes HTTe i l th Rn i dconsiderin advantagee th g f fueso l zoning, contro f coolano l t flow rat eacn i e h column, easy insertio f CRso n , irradiation flexibility in the core. Core components and reactor internals have been installed from May to August 1995. The installed core arrangement of a top replaceable reflector region is shown in Photo 2. In order to verify the seal performance between permanent reflector blocks, the air leakage test had been carried out to measure leakage flow rate e measure.Th d valu e less founb e wa s o t d thaassumee th n d limin i t core th e thermal hydraulic design e cor coole.Th s i e 4MPf heliuy o b d as flowinga m g downward. The core has 30 fuel columns and 7 CR guide columns, which is surrounded by replaceable reflector blocks.

The reactor cooling system is composed of the MCS, ACS and VCS as schematically shown in Fig. 1. The MCS is operated in normal operation condition to remove heat from the core and send it into the environment. The ACS and VCS have incorporated safety features. The ACS is initiated to operate in case of a reactor scram. Besides components sufficieno ha tw S f o VC t f tso ou capacite on removo t y e residual heate th , provides i S cooAC o t d l dowcore d cornth e an e support structure .helicallA y coiled intermediate heat exchanger (IHX) whose heat-resistant material is Hastelloy-XR develope JAERe bees th Iha y ndb installe Septemben i d r 1994. Nuclear heat application tests using the HTTR, are planned to be carried out, and accordingly a heat utilization system wil connectee b l e IHX e fueth .Th o l t d fabrication starte Junn di e 1995 and will complete in 1997.

e fueTh l fabrication facilit productioo tw s yha nkernee lineth f lso production and coating processes, and one fuel compact production line.

17 Phot . 2 oInstalle d core arrangemen p replaceablto f o t e reflector region

I PPWC : Primary pressurized I water cooler : Primar C G s circulatoP ga y I r SPWC : Secondary pressurized ] water cooler SGC S G C : Secondary gas circulator : Auxiliar X AH y heat exchanger AG C: Auxiliar circulatos ga y r

Pressurized Auxiliary water water pump air cooler u. Pressurized water air cooler

Auxiliary water pump Auxiliary Cooling System ' M»'n Cooling System

Fig. 1 Cooling system of HTTR

18 A functional test operation of the reactor cooling system will be performed from May 1996 to September 1997. Fuel will be loaded into the core around in September 1997 firse th td criticalitan expectes i y Decemben i d r 1997.

3. HTTR Testing Program for Advanced HTGR Development

e HTTTh R projec intendes i t establiso t d d upgradan h e technologth e y basis necessary for advanced HTGR developments. Based on the discussion about future prospects of advanced HTGRs for the preservation of the global environment and sustainable development typeo ,tw f futur so e promising HTGRs have been selectee b o t d developed. medium-sizee th s i e 1On ' d modula d severran e accident free HTG enablo t R e the siting in industrial complex with process heat utilization of hydrogen production, conversio f fossio n l fuels suc s coaa h l gasificatio r liquefactioo n n and/or gas turbine power generation. The other is a small innovative HTGR of a super safe and/or maintenance-free HTGR site isolaten i d d islandbasemene th r o sf som to e buildings. The design studies and R&D works on the innovative HTGRs have been started in orde mako t e besr f eth testino t g program usin e HTTRth g . Aroun 199n i de 8th results of these studies will be subject to the review of the Japanese Atomic Energy Commission revisio(JAECe th t ) a f long-terno m progra r researchmfo , developmend tan utilizatio nucleaf no r energy.

Some heat utilization system is planned to be connected to the HTTR and demonstrate formee th t r da secon e stagth f deo core presentt .A , steam-reforminf go methane is the first candidate. Also the HTGR with Gas-Turbine has been studied for assessment of technical viability.

Besides the demonstration of the heat utilization system, the JAERI plans to carry out safety demonstration tests to confirm the salient inherent safety features of the HTGR. In addition material and fuel irradiation tests as upgrading HTGR technologies wil conductee b l d after attaining rated power. Preliminary testn so selected research subjects suc s composita h e materiaC coateZr d dan l fuel developments, have been carriehigt a ht dou temperatur d undeean r irradiation.

3.1 Nuclear heat utilization tests The HTTR is designed to transfer the thermal energy of 10MW to secondary helium at temperature of 905°C and pressure of 4.1MPa through an intermediate heat exchanger (IHX).

Steam reformin f methano g r productiofo e f hydrogeno s adoptee i nHTT th Rr fo d heat utilization system, because of the following reasons.

1) Hydrogen n ideawoula e b ld energy carriee utilizatioth f i r n technologies converted into useful form f energo s y (heat, electricit d fuelsy an s wela )s a l e storageth thos r fo e , transpor safd an te handlin f hydrogego n would have been established successfully.

19 e demonstratioTh ) 2 coupline n steae th tes th f o tmf go reformin g HTTe systeth R o t m will contribute to the technical solution of all other hydrogen production system because basic system arrangement and endothermic chemical reactions for hydrogen production are similar.

) Stea3 m reforming system consist componentf so maturef so d technologie none th -n i s nuclear field.

) Stea4 m reformin economican a s i g l hydrogen production process commercializet a d presen non-nucleae th n i t r field.

Desig nHTTe workth R f steaso m reforming system have been started since 1990 aimin demonstratiot ga beginnine HTTth e n21se th t testh Ra tf f o t centurygo .

Design of main system arrangement, of key components such as a steam reformer and a steam generato f controo d an rl method, especially start-up procedurd an e development of computer code for transient thermal-hydraulics analysis have been conducted. The steam generator is designed to have a safety function as an interface.

In order to achieve hydrogen production performance competitive to that for fossil fueled plants, measures have been taken to enhance helium side heat transfer followine rateus o t , g bayonet typoptimizo t f tubee o d san e proces compositions sga .

orifice 1Th ) e baffles with wire nets attache outside th n steaf do eo m reformer tubes are installed.

2) The bayonet type of reformer tube is adopted to utilize the outlet process gas at the outlet of catalyst layer at the temperature of approximately 830°C. e temperaturTh ) 3 procesf lowerees o i s sga d from 450° 600°Co t C increaso ,t e eth heat input preventing carbon deposit. This design achieves heat inpu 3.6Mf o t W from helium gas to the steam reformer and 1.2MW from outlet process gas of catalyst layer to inlet process gas. Total heat input of 4.8MW are utilized to givo t eprocesd e steaan th hea s p mtu sga reforming process heat.

The main system arrangement is shown in Fig. 2. The high temperature secondary heliu s flowga m first sa t intsteae th o m reforme d theran n int superheateoa a d ran steam generator. The superheated steam is provided mainly to the steam reformer.

The heat utilization efficiency of 78% using steam reformer is possible to be achieved, competitive to non-nuclear process (80%-85%).

3.2 Assessment of HTGR with Gas-Turbine technologies Study Group of the HTGR with Gas-Turbine Power System (HTGR-GT) of the Japan Society of Mechanical Engineers has been set up since November 1993 to review and assess the HTGR-GT technologies and identify technical areas to be solved with their development. This study group has held a series of the meeting and will release a

20 Steam reforming system

, CO , 2 H

CH< HTTR (30MW) IsolatioI / . n 905°/ > C— « valve 600°C CataIyst tube Aux i I coo I e

t earS n r eformer (3.6MW) - Feed wa t e r

St earn Pressu r i zed generator r e I o co r e t a w (3.3MW) (20MW) Secondar Iiure h y n

Fig. 2 Steam reforming system connected to the HTTR report on its activities by the end of February 1996. During this period, voluntary survey and R&D works evaluating related technologies have been conducted by researchers and engineers of JAERI as well as of nuclear industries. Corresponding to these activities, JAERI is now planning to launch a new feasibility study on the HTGR-GT from 1996 to 2000, which is subject to appropriated funds.

3.3 Safety demonstration tests The following safety demonstration test e plannee HTTar s th o veritt Rn i d y inherent safety feature f HTGRso s .

1) Abnormal control rod withdrawal tests and 2) Coolant flow reduction tests

For these tests the code as a simulator has been developed for performing analyses of HTTR core transient accidentsd an s .

3.4 Material and fuel irradiation tests as upgrading HTGR technologies To develop advanced HTGRs (reactor outlet coolant temperature about 1100°C, power density 3-6w/cm and fuel burnup about lOOGWd/t are targeted respectively), 3 overall R&D program on upgrading HTGR technologies has been drafted in March 1995. However there still exists strong arguments among ourselves what type of advanced HTGRs should be mainly developed for the future. Long-Term program for research, development and utilization of nuclear energy revised by the JAEC in June 1994 says that since the technologies of HTGR with high temperature heat supply should be develope e broadef nucleao th e r us fo dr r energy e constructioth , s it e HTTd th Ran f no testing program shal e planneb l executed dan r establishinfo d upgradind gan e HTGth g R technology basis and to conduct various innovative and basic researches on high temperature technologies. This program does not directly address future deployment pla f advanceo n d HTGRs followin startee e HTTRw th g o e desigS .th d n studie d R&an sD works on advanced HTGRs in order to incorporate the results in the long-term program revised by the JAEC around in 1998.

At presen e firsth t t priorit s dedicatei y e medium-sizeth o t d d modula severd an r e accident free HTGR with high temperature heat. In order to upgrade core performance achieving higher power density, higher e burnu cord higheth an ef o p r allowable temperature desig n e coatefuel C limith Zr c compositf , dc/ o t d fuean l e materiar fo l contro sleeved ro l s have been pre-tested.

. 4 Concluding remarks

The HTTR is a high temperature gas cooled test reactor which has various aims and operational modes constructioe Th . e HTT s progresseth R ha f no d smoothl s firsit d tyan criticalit s foreseei y Decemben i n r 1997.

The various tests by the HTTR will make a great contribution to confirm salient characteristics of HTGRs including reliable supply of high temperature heat as high

22 as 950°C and high inherent safety and the application of high temperature heat from HTGRs to various fields will also contribute to relax global environmental problems. Furthermore, the HTTR has a unique and superior capability for carrying out high temperature irradiation tests not only for development of advanced HTGRs but also for basic researches such as new semi-conductors, super-conductors, composite material developments and tritium production and continuous recovery testing of fusion reactor blanket materials. The HTTR is highly expected to contribute so much to promoting the international cooperatio thesn i n e fields.

REFERENCES

1) Kunitomi, K. Maruyama, S. lyoku, T. and et al. 1994. Future prospects for advanced HTGRs-Survey of future promising HTGRs. Proceedings (In Preparation): Technical Committee Meeting on Development Status of modular HTGRs and their Future Role, Petten, The Netherlands. IAEA, Vienna, Austria. November

) Hada2 . 1994,K . Desig heaa f tno utilization systeconnectee b HTTRe o t mth .o t d JAERI-Conf 95-009, pp225-238. November

23 PROGRESS OF THE HTR-10 PROJECT

D ZHONG, Y XU Institute of Nuclear Energy Technology, Tsmghua University, Beijing, China

Abstract

This paper briefly introduces the main technical features and the design specification e HTR-1th f o s 0 Present statu d maian sn progrese licensth f o es applications e desig th ,d manufactur an n e maith f no e componente th d an s engineering experiments as well as the construction of the HTR-10 are summarised

. Genera1 l Introduction Backgroun1 1 d Considering the utilization of nuclear energy in next century, China has paid great attention to the development of advanced reactors which have good safety features, economic competitiveness and uranium resource availability The high temperature gas-cooled reactor was chosen as one of the advanced reactor types for future developmen d covere e Nationaan tth n i d l High Technolog ProgrammD yR& n i e Institute Th 198 6Nucleaf eo r Energy Technology (INET Tsinghuf o ) a University was appointed as the leading institute to organize and carry out the key technology development, the conceptial design and the feasibility study of HTGR in 1986- calleo s 1990 perio" e Sevent, dth f o d h Five- Year Planconceptuae Th " l design and the feasibility study report of the HTR-10 was completed in 1991 and then examine Expere th y db t Committe e Energth f eo y Technology e NationaAreth f ao l High Technology R&D Programme and approved by the State Science and Technology Commission (SSTC) Finally e 10MTh , W high temperature gas-cooled test reactor (HTR-10) project was approved by the State Council in March 1992 INET is responsible for design, license applications, construction and operation of this tes HTR-1e t th reacto w beins 0i No r g constructe site f INEth eo n dTi whics hi located in the North-west of Beijing city and has erected other two test reactors, e g the 2MW swimming pool type experimental reactor and the 5MW nuclear heating test reactor

1 2 The Objective of the HTR-10 e constructioTh HTR-1e th firse f th o e t n HTGs 0 i th ste f po R development strategy in China The objective of the HTR-10 is to verify and demonstrate the technigue and safety features of Modular HTGR and to establish an experimental bas r developinfo e nucleae gth r process heat application e specifiTh s c aims of the HTR-10 have been defined as follows •' (1) To acquire the experience of HTGR design, construction and operation (2) To carry out the irradiation tests for fuel elements (3) To verify the inherent safety features of the Modular HTGR (4) To demontrate the electricity/ heat co-generation and gas/steam turbine combined cycle develoo (5T ) hige pth h temperature process heat utilizations

25 Desige 1.Th 3 n Specification The HTR-10 is a pebble bed type high temperature gas-cooled reactor, it uses the spherical fuel elements with ceramic coated fuel particles. e reactoTh r core whics hha diamete f 1.8o r , mean heigh volume f 1.97o tth d m5.0mf ean o surroundes i 3 e th y b d graphite reflectors. 27,000 fuel elements are loaded in the core. The fuel elements use enrichmenw thlo e t e uraniudesigth d nm an mea n burnu s 80,00i p 0 MWd/te Th . pressur primare th f eo y helium circui . OMPafirse 3 th s i t n tphase I . e HTR-1th , 0 wil operatee b l core th et da outle inled an tt temperaturtemperatur t " 0 70 f o f eeo 2501. At the secondary circuit, a steam turbine cycle for electricity and heat co- generatio s designedi n e steaTh m. generator produce e steath s t temperaturma e of440t and pressure of 4.OMPa to provide a standard turbine-generator unit. The main desig nHTR-1e datth f eshowo e 0ar Tabln ni . e1

Table 1. The main design data of the HTR-10 Reator thermal power 10MW Primary helium pressure 3.OMPa Core outlet temperature 700t Core inlet temperature 250

In the second phase, the HTR-10 will be operated at the core outlet temperature of900 inled an t temperaturturbins ga °A c0 e. (GT30 stead f eo )an m turbine (ST) combined cycle for electricity generation is preliminarily designed. The intermediate heat exchanger (IHX) with therml power of 5MW provides the high temperature cycleT steae G nitroge Th e 850mf . o th r s generato1nfo ga r (SG) with rest thermal power of 5MW produces the steam at temperature of 4351: for the ST cycle. The main design data of the second operation phase are shown in Table2.

Table 2. The main design data of GT-ST Combined Cycle • Reactor Thermal power 10MW Primary helium pressure 3.OMPa Core outlet temp. 900

26 • Power output Power of GT cycle 1.94MWe PowecyclT S f er o 1.36MWe

1.4 The Main Technical Features Comparing wit previoue hth s pebbl HTGRd ebe e particulath , r technical features of HTR-10 are as foil owing: e cor(1Th )e residual hea designes i t dissipatee b o dt passiva y db e heat transfer system. (2) The two reactor shut down systems which are consisted of 10 control rods an 7 smald l absorber bal l lpositione holeal e e sidsar th e n i reflectord corn i e e Th . control rods are not needed. e pneumati(3Th ) c pulse singulize r discharginrfo sphericae gth l fuel elements from reactor cor fue e uses ei th l n dhandlini g system. reactosteae e th (4Th md ) ran generato arrangee ar r d sidpressur e sidey eb Th . e boundary of the primary circuit is consisted by the reactor vessel, the steam generator vessel and the connected vessel (hot gas duct vessel) e integrate(5Th ) d steam generato intermediatd an r e heat exchangee ar r designed. The SG is a once through, modular small helical tube type. The IHX can provid hige eth h temperature nitroge heliur no f m850o o secondar900

2. Progress of the HTR-10 Project 2.1 The Safety Review and Licensing Application applicatioe th r constructioe Fo th f no n permission followine th , g procedures should be passed. (1) INE compiled Tha Environmentae dth l Impact Report (EIRHTR-1e th f )o 0 and submitte Nationae th o t t di l Euvironmental Protection Administratio n( NEPA ) f 1992o repord e ,revieweth s mi e wa ti nth expera y db t committee, thee NEPth n A HTR-1e th f o Novembe n 0e i necessar approveth R EI f o e re yth d1992 on s i t .I basis for the application of the reactor site. (2) The Siting and Seismic Report (SSR) of the HTR-10 was submitted to the National Nuclear Safety Administration (NNSA). After examinatio e reactonth r site was approved in December 1992. (3) The Preliminary Safety Analysis Report (PSAR) had been completed and submitte e e applicatioNNS th th e constructio r th o At fo f d o n n permission i n December 1993 activitiee Th e .licensin th f o s g procedur e eyear on laste .r fo d The NNSA formally issued the construction permission of the HTR-10 in December 1994.

2.2 The Design of the HTR-10 desige th licensin d r nan (1Fo ) g reguirement INEe th ,prepared Tha technicae th d l document sdesig e whicth e nhar criteri HTR-1 e e formath th f d ad o conten 0an an t t safete th f yo analysis HTR-10 e reporth f o t . Thes o documenttw e s were examined and approve NNSe th y AugusAb dn i t 199 d Marc2an h 1993 respectively.

27 (2) The basic design and the budget estimate of the HTR-10 was carried out in the mid of 1994 and then examined and approved by both the State Education Commissio ne Stat(SECth d e an )Scienc d Technologan e y Commission (SSTCn i ) f 1994o d th.een (3) The detailed design of the components, systems and buildings is being carried out by the INET in Cooperation with the Southwest Center of Reactor Engineering Researc Desigd han n e (SWCRheliuth r mfo ) purification syste d othean m r helium auxiliary systems, and the Beijing Institute of Nuclear Engineering (BINE)© for the steam electricity conversion system and the turbine generator building. For the detailed design of the main components e.g. the reactor pressure vessel, the steam generator and the helium circulator, the design engineers of the INET have close contacted and discussed with the engineers of the manufacturers to modify and improv e desigth e e technican th drawin d an l gspecifications e detaileTh . d design of the HTR-10 is planed to complete in the next year.

2.3 The Engineering Experiments A engineering experiments program for the HTR-10 key technique has been performe INEn i d yearsr Tmaie fo Th . n e engineerinaimth f so g experimento t e sar verify the design characteristics of the components and systems, to demonstrate the relevant features and to obtain the operation experience in the simulated conditions. The various experimental facilities have been set and tested or are being established. The key engineering experiments are as following: • The high temperature helium test loop and the relevant helium technology. fuee Th l •handlin g system test. controe Th drivind • ro l g apparatus test. smale Th •l absorber ball simulating system. s ducga tt tesho te facilityTh • . • The stability test of the steam generator model. e heliuTh • m flow temperature mixing. e tesTh t component fuee th l f handlino s ge smalsysteth d l man absorbe r ball system prototype th ,coutro e th drivinf d o ero l g apparatu tese th t d sectioe san th f no hot gas duct are designed in 1: 1 scale. It is planed to perform the experiments at the operation temperature and helium atmosphere conditions. The experiments of the fuel handling syste d smalan m l absorber ball simulating syste t a mambien t temperatur beed eha n carried out.

Manufacture Th 4 Maie 2. th f neo Components e maiTh n component HTR-1e th f o s 0reactoe sucth s ha r pressure vessele th , steam generator vessel and its internal parts, the helium circulator and the core matallic interna fabricatee ar le domesti th y b d c factories which hav e abiliteth d yan experience of manufactring PWR's components. The graphite of the core internal and safete parth f yo t grade helium valves e wilimporteb l d froforeige mth n suppliers. The reactor pressure vessel is a safety grade I component. It is a cylindrical vessel whic heighs hha f 11.4mo t , diamete 4.2f d totaro man l weigh f 142tonso t s i t I . fabricated by the Shanghai Boiler works. e steaTh m generator vesse pressure th pars a lf o t e boundar primare th f o yy circuit is also a safety grade I component. It has height of 11.2m, diameter of 2.5m and total weight of 70tons. The once through type steam generator is consisted by 30 small helical heating tubes. The diameter of the heating tube is 18x2mm/18x3mm and the effective length is 35m. The helical tube unit is 115mm

28 in diameter. This component (vesse d internalsan l s fabricatei ) e Shanghath y b d i Power Station Auxiliary Equipment Works. The helium circulator is a vertical single-stage centrifugal one, the impeller is at shafte circulatoe th f samTh e .o th d ths e een axlrha e wit drivs hit e moto fixed an rn di the circulator pressure e vessesteath parp f mo lto twhic e generatoth s hi r vessel. e heliuTh m circulato fabricates ri Shanghae th y db i Blower Works. The core metallic internal which consistes of the metallic core vessel, biological shielding structure, bottom support plate and enhanced plate, top pressing plate will manufacturee b Shanghae th Machiny 1 db . iNo e Works. e graphitTh e brick corf so e reflector wil suppliee e b lToy th oy d b Tanc . LtdCo o. Japan finae th , l machinin worke planes gi done th b f INET n o so ei dt e carbo Th . n brickreflectoe th f so r wil domesticalle b l y fabricated. The components of the fuel handling system, the helium purification system and other auxiliary systems will be also domestically made.

Buildine Th 5 g2. Construction e HTR-1Th 0 test plant include reactosa r building turbina , e generator building with two cooling towers and a ventilation centre with a stack. The buildings are to be arranged and constructed in the area of lOOx 130m2. The building construction and installation are contracted by the Engineering Company No.23 of the China National Nuclear Corporation (CNNC) The excavation ground was completed in the end of 1994, the constraction of the HTR-10 was formally started in June 1995, the basement of the reactor building was poured in September 1995.

3. The Time Schedule

Tabl . Time3 e Schedul HTR-1r efo 0

1994 1995 1996 1997 1998 1999

1) 11 1 Milestone Construction First Concrete FSAR,Critical Power .icence Operabnon Operatior

Basic Design Detailed Design Construction • Site Preparation • Building • Manufacture of Components • Installation Commissioning • Critical • Test • Power Operation

The time schedule of HTR-10 construction is shown in TableS. The reactor building construction will be lasted two and half year and scheduled to complete in the end of 1997. In parallel, the manufacture of the components and installation of the main components, equipment auxiliard san y systems wilclose b l e followe progrese dth f so e buildinth g constructio d schedulean n f 1998 o e firs d o completTh t .ten d e th n i e criticalit reactoe th f vo s plane i r reachee b besmnine o dt th n di f 199ao 9

29 STATE OF HTGR DEVELOPMENT IN RUSSIA XA9642776

V.F. GOLOVKO, A.I. KIRYUSfflN, N.G. KODOCfflGOV, N.G. KUZAVKOV OKB Mechanical Engineering, Nizhny Novgorod, Russian Federation

Abstract

Detailed design f VGR50so M reactor, VG VG40 d 0an s were developed / in Russia e mos.Th t important e activitresulth f o ts formatio ywa f o n contacts between different organization d creatioan s f o technologn y basis for HTGRs. At present it is assumed that advantages of HTGRs as compared to other types of reactors can be more completely demonstrated if module reactors with thermal power of about 200 MW and pebble bed core will be aimed only at process heat production, and for electricity production through gas-turbine cycle will be used module reactors with therma a cor d le an powefro W M mabouf r o 0 prismati60 t c blocks. To this effect, feasibility study of VGMP reactor was carried developmend an t ou f GT-MHo t R with gas-turbine cycl undes i e r way.

1.Brief information on HTGR development history in Russia.

Development of HTGRs was initiated in Russia about 30 years ago. During this period design VGR-50f so d ,VGM an VG-400P M reactoVG . r plant have been worked out. General information on the projects is given in Table 1.

heliu- VG0 5 R m cooled reactor with pebbl d cors ebe intendee wa r fo d electricity production and radiation of polyethylene tubes. To this effect circulatio f sphericano l fuel elements aroun a closed d path was provided.

31 TABLE 1.

Name of the Dates of beginning General Designer Phase of the project and completion of of the reactor development the project

VG0 5 R 1963-1985 Research Institute detailed of Machinery for design Atomic Industry, Moscow VG 400 1974-1987 OKBM detailed Nizhny Novgorod design VGM 1986-1991 OKBM detailed design

VGMP 1991-1992 OKBM feasibi1ity study

Parameter reacto0 5 R e showVG rar Tabl n f si o n . 2 e

TABL. E2

Parameter Magnitude

Therma1 powert ,MW 136 Electrical power, MWt 50 Helium temperature, °C reactor input 260 reactor output 810 Helium flow rate, kg/s 54 Helium pressurea .MP 4

32 a numbe r reasonf Fo ro s activit e desige th th s ende n n nwa o yi d middle of 80-th.

VG 400 reactor plant was intended for both electricity production and process heat e hea transferre.s Th i t methana o t d e steam reformer * and hydrogen gained is used for ammonia production. The reactor plant parameter y component ke e give e e sar reactoTablth n Th i n . f 3 eo s r plant are housed in the prestressed concrete reactor vessel (Fig 1).

Parameter VG40f so 0 reacto showe ar rTabln i n 3 e TABLE3.______

Parameter Magnitude

Thermal power. MWt 1060 Electrical power, MWt 300 Number of loops 4 Helium temperatureC .° 0 35 reactor inlet 0 95 reactor outlet Helium temperatur steat ea m generator inlet. °C 750 0 34 Helium flow rate, kg/s « 5 Helium pressure a ,MP

On the preliminary phase of the design development two variants of the core were analyzed: on the basis of pebble bed and prismatic fuel blocks resulu s f A o calculational.t , desig d engineerinan n g analysis the pebble bed core was chosen for further development. The pebble bed core option for this reactor was made taking account of the following considerations:

33 Fig.l. VG-400 reactor plant

- stea1 m turbine plan6 t PCRV 2- intermediate circui7 t steam generator 8 helium gas circulator - 3 fuel loading syste9 m bypass valve 0 1 drived ro S s CP - 4 fuel unloading system - 1 relie5 1 f valve intermediate heat exchanger

34 more simple technolog f o fuey l elements manufacturd an e possibilit f theio y r full scale testin n experimentai g l reactors; - use of more simple mechanisms for the core refueling; possibilit- e corth e f refuelino y g witreactoe th h r running. 0 reactoV40 G r with closed gas-turbine cycle ,was analyzes a d well.

VGM modular reactor was designed to validate main technical / decisions connected with productio f higo n h temperature process heat. VGM reactor operation experience would promote later utilizatiof o n HTGRs in the Russian Industry. On the basis of VGM reactor the development of VGMP reactor for production of middle temperature heat was initiated. Main parameters of the reactor plants are given in Table 4.

TABLE 4

__Magnitude_ VGM VGMP

5 21 0 20 Thermal power t .MW Helium temperature. C reactor inlet 300 300 reactor outlet 750...950 750 Helium flow rate, kg/s 59...85 91,5 6 5 Helium pressure a ,MP Number of loops 1 main 1 main and 1 auxiliary

e reactoTh r component e arrangear s n steei d l vessels (Fig2 and 3).

35 - 1reacto r 7 - fuel circulation system 2 - pressure vessels unit 8 - small absorber bails system - 3intermediat e heal exchange - heliu 9 mr purification system 4 - steam generator 10 - relief valve circulatos ga - 5 r 11 - steam-turbine plant 6 - surfase cooling system

Fig. 2. VGM reactor plant

36 Intermediate circuit helium

L_ _ J

Fig 3 VGM-P RP schematic diagram 1 - reactor core, 2 - vessels block - cor 3 ,e protective housing - intermediat 4 , e heat exchanger, s blowega m r witma a cut-ofh- 5 f valv- relie 6 oe f valve - heliu 7 , m purification system - fue 8 , l circulation ^vsie - smal 9 m l absorbing balls system - fue 0 1 l. element discharg d drivero S . e CP facility - 1 1 , 1- 2loadin g volum f - SABSemergenc o e3 1 , y cooldown system 14 - localized valve

37 . Stat2 HTGRf o e s component development.

The great attention was paid to the development of high temperature heat exchanger, steam generator, circulator, fuel, graphit d structuraan e l materials. Different designe f o sstea m generator and heat exchanger have been considered, experimental facilitie e investigatioth r fo s f o nthermohydrauli d vibratioan c n

characteristics have been created (ST-1312, ST-1565).The 5 MWt /steam generator model has been fabricated to be tested at ST-1312 facility e facilit trian Th .i s l wa yoperatio heatere th t na s powef o r e nominath f 3l0%o value e heliuTh . m temperatur pressurd an e e wer0 73 e °C and 5 MPa. A full-scale prototype of the gas circulator has been fabricated r testinfo t ST-138a g 3 facility. Startu d triapan l operatioe th f o n facility under partial loads have been performed. The further testing s beeha n delayed. Russian organization has considerable experimental background in coated particl d fuean e l compact manufacture A .pilo t plant witn a h

Q output of 10 coated particles per year was developed The Chemical Concentrates Plant (Novosibirsk s procesha ) s equipmeno t manufacture fuel particles and fuel elements. The manufacturing process employ e "sol-gelsth " typee plans Th produce.ha t d about 10,000 spherical fuel elements e s fueTh irradiateo .wa t l p u d temperature 1600°Cf so . The Research Institute of Graphite (Moscow) has been developed technolog RBMf o y K graphitw GR-1ne G d graphite.MP an e s whice ar h based on calcinated and noncalcinated coke pressing. Test specimen G graphite GR-f MP so d 1an s were irradiated under fluencies in the range of 0,5 10 22 ...2,5 10 22 cm —2 and temperatures in the range of 500 °C...1000 °C.

38 The development of structural materials for reactor components was e Researcth carriet a ht ou Institutd e "Prometey"(S.Petersburg). Perlite type steel for reactor vessels is certificated and can be used up . C Materia 0 35 t of stea o l m generator tube stainless i s s steel HC-33. The characteristics of the steel are validated at test.facilities under temperatures up to 750 °C The most complex operating conditions are concerned to heat exchanger. Chrome-nickel alloy *lC-57 is a material for heat exchanger tubes strengts It . validates i h t temperatura d f o e 930°C during 10000 hours.

3. Assessemen f statuto d prospectsan s of HTGRs development in Russia. e informatioTh n presented shows than Russii t a cosiderable experienc HTGRf o e s developmen s beeha t n accumulated. The long-term activity on the HTGRs projects led to apperance of cooperation between various organizations capabll al o t carrt e ou y necessary scop researchf o e , developmen d desigan t n works. There ar e sufficient number of facilities for testing of HTGRs equipment, irradiation of fuel, graphite and structure materials. However, of late years activities associated with the development of HTGR project n Russii ss sharplwa a y o t reduceeconomi e du d c reasons. What are prospects of HTGRs development in Russia under existing conditions? There is weighty background for the statement that HTGRs should play significant role in nuclear energetics of Russia. The objective need for HTGRs is clearly defined toy their unique potential. Indeed, temperature level to 350°C is an area of LWRs application, temperature leve 550°o t l C belong o LMRs,uppet s r o 1000°levet p s u lCi within the capacity only of HTGRs. Therefore, the task is efficiently to use the advantage of HTGRs. The following ways are preferable for that.

39 The first is production of high temperature heat for various industrial processe knows i t nI s well that consumptio f o fossin l fuelr heatinfo s d industriaan g l processe -mors i s e r thafo n electricity production. As far as demand in energy on fossil fuels face growinl al s g ecological, reserv d transporean t problemss i t i , inevitable more broad applicatio f o nnuclea r plant s a alternativs e source energyf so . Basic consumption of heat in industry falls on temperatures up to 550-60 . Thi0C sprovidee b leve n ca ly HTGR b d s with helium temperature of 750°C. l Almosrefinerieoi l al t d plantan sr productiofo s f o n petro d diesean l l fuel fro e servemb coan HTGRy ca b dl s with suca h level of temperature This allows to save only for the oil refineries processedl oi e th f .o abou % 15 t It should lay emphasis on additional unique feature of HTGRs which allow orientato t s e them towards producing process heat, namely capability of module concept of HTGRs to provide passive removal of residual heat through the reactor vessel surface by natural processes (convection, conductivity, radiation). Moreover, unlike other typef so reactors the residual heat is removed not only under loss of flow but e coolanith f t escaped froe primarth m y circuit additionn I . , maximum fuel temperature doesn't exceed design basis limi f o 1600°Ct f i . nominal reactoe poweth f o rr withd pebblan d corW M aboube s e i e0 20 t with prismatithin I s. MW casaboue possibilit e 0 on cth e 60 t f o ycor e meltdown and its relocation is completely excluded taking into account very high temperatur graphitf eo e sublimatio >3000°C( n ). This enable o assurt s e high leve f safeto o t locatl d an ye HTGRs r fronofa tm process plants wha s i necessart y from economic point of view.

If projectM VG-40VG d 0an s were aime t combinea d d productiof o n process head electricityan t , VGMP projec s completeltwa y orientated towards productio procesf o n s heat. VGMa s desinestandar wa Pr fo d d

40 oil-refinery plant with taking accounUser'e th f so t requirementse .Th feasibility study showed that VGMP can provide process heat supply for all basic e regimeoil-refinerth f o s y plant e nucleaTh . r plant includes three VGMP reactors. The heat is transferred through intermediate circuits to a process one (Fig 4).

NPPS ORP •///——»SOOT - t.O ftPc r —H

HPa 6.0

-I/

flPo1.0 SOOt

6.0 flPa

soar t.o

L_ _ _".______) -/- - Primary circuit -II- - Intermediate circuit - Heating-III- system

Fig.4. NPPS flow scheme: 1 - reactor; 2 - intermediate heat exchanger; 3- mai blowers nga - intermediat 4 ; e circui blowers ga t proces- 5 ; s heat exchanger; 6 - technologic, circulator

41 M reactoVG backgrouna s i r r developmenfo d f VGMPo t . what enables e scientifitus o d technicaan c l potentia s beeha l n o t create d an d elaborate the design if necessary, for short period. The second way is use of HTGRs only for electricity production. e noteb o dt s howeverIi t , that connectio HTGRf o n s with conventional steam-water cycle doesn't correspon o theit d r capabilitie d can'an s t provide competitivenes f HTGRo s s with other source f o energys . Realisatio f HTGo n R advantages shoul da combinatio loor fo k f o nHTGR s / with new technologies. For that matter, a coupling of HTGRs with gas - turbine n advancecycla s i e d solution. Thus, at present in Russia changes in strategy of HTGRs utilasatio s formedha n , namely: r productiofo - onlf no y process heat, using module typf o e HTGRs with therma pebbld l an powed core W be eM abouf o r,0 instea20 t d of combined function.- high temperature part (700°C-950°C r proces)fo s heat, low temperature part (300°C-700 C) for electricity production through steam-water cycle; r electricitfo - y production through gasturbine cycle using module typ HTGRf o e s with thermal powe f ro abou ta cor 600Md e an W formed from prismatic blocks. The strategy is assumed to more completely demonstrate the advantage f thiso s typ f reactor-so e .

. Activitie4 s relate HTGRo t d s wit s turbinhga e cycle.'

Initial work on direct gas turbine cycle was performed at OKBM in early of 80-th as applied to VG 400 reactor. The layout and parameters of the reactor are given in Fig.5. The cavities of the reactor vessel are contained the core, two horizontal turbomachine, recuperators and coolers. Residual heat removal under normal and accident conditions is

42 to cooling tower 75 °c Shutdown circulator /Core PCRV ,for heeting 170°C in

Heat exchanger ' 1 ecuP^ erator

Generator N=200MW

Compressor purification system ^Shutdown heat exchanger

Fig.5 VG 400 with gas turbine cycle Fig. 6. GT-MHR reactor arrangement 1 - reactor core; 2 - core outlet plenum; —3 intercooler - precooler 4 ; ; - 5compressor - turbine 6 ; ; 7 — recuperator - generator 8 ; ; 9 — core inlet plenum.

44 provide o independentw y b d t loops with circulator d coolersan s e Th . works on the project couedn't be continued dne to interruptio designinf o n 0 reactor40 G V g . GT-MHR project is a joint Russian-American design of the nuclear power plant with direc s turbinga t e cycle. GT-MHR projec beins i t g develope e grouth f o Russia py b d n enterprises heade OKBy b d M (Nizhny Novgorod). General Atomics (San Diego e )leadin th wile b gl firm from American side. Russian GT-MHR project is based on the concept of General Atomics reactorll]. GT-MHR plant is a coupling of passively safe modular reactor with helium coolant with up-to-date technology developments: large industrial gas turbines, large active magnetic bearings, compact highly-effective heat exchangers, high-strength high temperature steel alloy vessels. Fig.6 show commoe sth n genera lreactore th vie f o w . The three vessels are located in the underground silo. e reactoTh r vessel house e annulath s r reactor core, core supports, control rods drives, heat exchange s circulatoga d ran r of the auxiliary loop. The reactor core contains hexagonal graphite fuel columns, which contain fuel (weapon- grade plutonium) encapsulated in ceramics coated particles. The reactor vessel is surrounded by a reactor cavity cooling system, which provides totally passive decay heat removalA . separate cooling system provides decay heat remova refuellinr fo l g activities. e poweTh r conversion vessel contain turbomashine th s d an e three compact heat exchangers e turboraashinTh . e consista f o s generator, turbine and two compressor sections, mounted on a single shaft suspende magnetiy b d c bearings. • GT-MHR plant has the following parameters: Reactor power 550-600 MW(t) Turbine inlet temperature 850°C Reactor inlet temperature 490°C Compressor inlet temperature 26°C Turbine inlet pressure 7,02 MPa Overall pressure ratio 2,86 Primary circuit pressure losses 7-8 * By-pass helium flow rate for cooling linga se s e d leakth an n i s 2.5%

45 Recuperator efficiency 0.95 Generator rotational frequency 50 Hz Turbine adiabati% 93 c efficiency w pressurLo e compressor adiabatic * efficiency 88% High pressure compressor adiabatic efficiency 87% Generator efficiency 98,5% Plant efficiency 47-48t %ne At presen e conceptuath t l desig f GT-MHo n beins i R g carrved out and capabilities for performing tests of fuel and main components are being analyzed.

5. CONLCUSIONS 5.1. Russian organizations accumulated significant experience associated with HTGRs developmen e basith f o sn o t M projectsVG VG40 d an 0 . Experimental facilitie r testinfo s f o g main component reactore th f so s have been created. At presen e strategth t f o HTGRy s utilasatio s beeha n n changed. Module HTGRs with pebble bed core will be aimed only at process heat production, and with a core from prismatic blocks at electricity production through gas-turbine cycle r Ro thi. s purpos a efeasibilit y stud f VGMo y P reacto s beerha n perfomed an d conceptual desid f GT-MHno s undei R r way.

REFERENCES

] Simo[1 n W.A. etal. "Desig s TurbinGa n Featuree th e f o s Modular Helium Reactor (GT-MHR)". 4th Annulr Scientific and Technical Conferenc e Nucleath f o er Society. 1993. Nizhny Novgorod. Russia.

46 HIGH TEMPERATURE REACTOR DEVELOPMENT XA9642777 IN THE NETHERLANDS

HEEA.IN .VA K Netherlands Energy Research Foundation ECN, Petten, Netherlands

Abstract This year, some clear design choices have been WHITEthe made in Reactor development programme. The activities will be concentrated at the development of a small combinedsizefor pebbleHTR heatpowerbed and production with closeda cycle turbine.gas Objective developmentthe of threefold:is restoring1. social support, 2. commercial viability after market introduction, and 3. to make the market introduction itself feasible, i.e. limited development and first-of-a-kind costs. This design is based on the peu-a-peu design of KFA Julich and will be optimized. The computer codes necessary thisbeingfor are prepared thisfor work. dynamicThe neutronics code PANTHER is being coupled to the thermal hydraulics code THER- MIX.-DIREKT. thisFor reactor type, fuel temperatures maximal are scenario the in of depressurization with recriticality. Even for this scenario, fuel temperatures of the 20MWth PAP-GT exceednot do 130CPC, thereso should roombe upscalingfor for economic reasons.otherthe On hand, wouldit convenientbe fuelreactorto the batchwise instead of continuously, and the use of thorium could be required. These features two a largerlead may to temperature margin. optimalThe design must unite these features in the best acceptable way. To gain expertise in calculations on gas cooled graphite moderated reactors, benchmark calculations beingare performed parallelin with international partners. Parallel to this, special expertise is being built up on HTR fuel and HTR reactor vessels.

1. Introduction R&D activities dedicated to the helium cooled graphite moderated high temperature reactor (HTR) started in 1993. In 1994 the name WHITE Reactor was given to the programme, standing for Widely applicable High TEmperature Reactor. The programme is mainly executed by ECN, although three partners contribute in the framewor e Programmth f o k Intensifo t e e Nucleath y r Competence (PINKe th f o ) Dutch Ministr f Economiyo c Affairs . Interfaculte Thesth e ear y Reactor Institutf o e Delft, the engineering company Stork NUCON and the utility research institute and engineering company KEMA. Parallel to this, an HTR Technology Assessment Study is being performed by ECN and the University of Utrecht. This study mainly investigates societal aspect technologye th f so . In 1994hosteN EC n IAE, a d A Technical Committee Meetin n Developmeno g t Status of Modular High Temperature Reactors and their Future Role, and the Dutch activitiewerR eHT presenten o s Workshoa n do p organize occasioe th e n th d o f no TCM [1].

2. Design requirements and basis

The objective of the WHITE Reactor programme is to contribute to HTR developmen r safefo t , environmentally friendl d economicaan y l energy supply. Three designR HT e : th r fo t requirementme e b o t e sar

47 1. restoring social support, 2. commercial viability after market introduction, and mako t e marketh e . t3 introduction itself feasible, i.e. limited developmend an t first-of-a-kind costs.

To comply with these requirements, some clear design choices have been made in the WHITE Reactor development programme this year. Design efforts are focused on simplicity. To restore public support for nuclear power, designs will have to be very transparent emergenco N . y core cooling systems, backup shutdown systemr o s containments shoul needede db . Thi sr leve r unitlimitpe pe modularizatioo le s , sth n will be needed to comply with larger power level demands. More important, the low power level positiva eve s nha e effec plann o t t economics, because componentn ca s be omitted and the safety grade of certain components could be lower than usual for nuclear power plants. A small plant size will also benefit market introduction, because buildin "adventuren smala ga e b l unin ca f limitet o " d sizee activitieTh . s will therefor e concentrateb e developmene th t da smala f o tl sizR e pebblHT d be e for combined heapowed an t r production wit hclosea d cycl turbines ega powee Th . r leve moro l n wil e eb l thaMWth0 n10 keeo t ,desig e pth s simplna s possiblea d ean to minimize development and prototype cost. Pebble bed fuel is chosen for two reasons: 1. the possibility of continuous fuelling, and therefore assure a limited overreactivity . les2 sd developmenan , t work remaine pebbl done th b n o eo t s fuel type than on the prismatic fuel type. The combined heat-and-power application has been chosen becaus t e chosefitei th s n powe rcove o t rang d a markerean t segment additional to large scale base load electricity generation for the nuclear industry. The design is based on the GHR-20 design by BBC/HRB [2] and on the peu-a-peu design of KFA Julich [3]. This is an extremely simple pebble bed HTR design. The reactor is being fuelled on-line continuously or in small batches. After a period of several years e reactoth , s unloadei r d off-line. Lik l modulaal e r HTRs e reactoth , r employ emergenco n s y core cooling system, since even afteworse th r t core heatup accident (depressurization) the reactor is shut down by it's negative temperature coefficient and the fuel does not attain unacceptably high temperatures, offering ample time for an active shutdown by insertion of rods. But the shut-down by negative temperature coefficient is only temporary, because of cooling of the fuel and decay of xenon. To decrease the safety function of the shutdown system, the reacto s alsi r o require o copt d e with recriticality after core heatup. Calculations performe JulicA KF h t a dsho w thi s possibli s e with0 4 wid d e an margin 0 2 r fo s MWth peu-a-peu designs. It is intended to uprate the power level and to optimize the design.

3. Computer code corr sfo e design

Although the pebble bed neutronics/thermal hydraulics code VSOP of KFA Jiilich is available through the NEA Databank now, ECN decided to develop it's own code system. The dynamic neutronics code PANTHER has been acquired from AEA, UK. This cod s beini e g couplee thermath o t d l hydraulics code THERMIX-DIREKT, kindly delivered to ECN by KFA Julich in a cooperation framework. Much attention is being paid to the generation of nuclear data. For the PANTHER- THERMIX system, neutron cross section generatee ar s WIMS-y db SCALE-d Ean 4 codes e Monte-CarlTh . o code MCN uses P i checo r thi t s well dfo a sd kan , certain reactor calculations.

48 To gain expertise in calculations on gas cooled graphite moderated reactors, benchmark calculations are being performed with these codes in parallel with international partner n similao s r systems e PROTEUth : S experimen I Villigene PS th t d a t an , 450MWth MHTGR of General Atomics.

4. Fue pressurd an l e vessel

Paralle thiso t l , special expertis reactoR HT fued rbeins R e i ves an l HT g -n builo p u t sels. Experiments are being performed on oxidation and interaction with fission e stabilit th s wela products a lC f UO yo Si 2f /UCo s 2 mixtures e pressurTh . e vessel stress assessment code ISAAC is being equipped with a fracture mechanics and creep assessment feature for high temperatures, to be able to design a licensable pressurR HT e vesseneae th r n i lfuture .

5. Future plans

In 1996, static and dynamic calculations are being performed on the 20MWth PAP- JiilicA compared KF hGan f To theio dt r calculations with different codes r thiFo .s reactor type, fuel temperature e maxima ar sscenarie th n i l f depressurizatiooo n with recriticality. Even for this scenario, fuel temperatures do not exceed 1300°C, so there should be room for upscaling for economic reasons. On the other hand, it would be convenien o fuee reactot t th l r batchwise insteaf o f continuouslyo e d us e th d an , thorium could be required. These two features may lead to a larger temperature margin. The optimal design must unite these features in the best acceptable way. An economic evaluation of the optimized design will be made. Attention will be paid to the possibility of omitting components and decreasing of the safety grade of components.

5. Conclusion

After a familiarization phase, the Dutch HTR program is well under way. First decisions have been mad design eo conceptua R nHT featuree th r l e fo desigs Th n ni Netherlands e activitieTh . s wil concentratee b l developmene th t da smala f o t l size r combinefo R pebblHT d d heapowed be ean t r production wit closeha d cycls ga e turbine. An independent neutronics and thermal hydraulics code system is being developed optimizen A . d design GHR-2e basePAP-Ge th th n d o 0an s Tplanne i d for 1996.

REFERENCES [1] Proceedings of the ECN Workshop on the Role of Modular HTRs in the Netherlands, Petten e NetherlandsTh , 0 Novembe3 , Decembe1 - r r 1995, ECN-R-95-027. s coolega dW heatinM . Schmitt0 H 2 [2 e g] reactoTh , r upgradin GHR-10e gth , Nuc. Eng. Des(19901 12 . ) pp.287-291.

[3] E. Teuchert, K.A. Haas, Features of safety and simplicity of advanced pebble bed HTR's, Proceedings of the IAEA Technical Committee Meeting on Development Status of Modular High Temperature Reactors and their Future Role, Petten, The Netherlands, 28-30 November 1994, ECN-R--95-026.

49 ERENCH ACIWmES ON GAS COOLED REACTORS

D. BASTIEN DMT/Dir, CEA/SACLAY, Cedex, France

Abstract

Thcooles ega d reactor programm Francn ei e originally consiste f eighdo t Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantlin f decommissionego d facilities, includin developmene gth f methodo t r sfo dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R&D programme. France actively participates in three CRPs set up by the IAEA.

1NATURA- L URANIUM GRAPHITE-GA- S COOLED REACTORS

France has built and operated 8 Natural Uranium-Graphite-Gas cooled Reactors (UNGG exportated an ) Spain i e ndon whic operates hwa Frencha y db - Spanish company (Hifrensa). The two first one, Marcoule G2 and G3, was operated by CEA and later by COGEMA. The other, Chinon 1, 2 and 3, Saint Laurent 1 and 2, Bugey 1, was operate EOFy db . Toda thef o shutdown e l myar al table .Th givee 1 s dat r eacafo f ho them.

studiee Th s relate thesdo t e reactors concern monitorin dismantlind gan f go decommissioned facilities. The CHINO reacto1 NA convertes wa r d int nucleaoa r museum whic verhs i y much visited. The other gas cooled reactors are being dismantled. The MARCOULE G2 and G3 and CHINON A2 reactors have been dismantled to level 2. CHINON A3 have been dismantled to level 1 and is waiting for administrative authorization to undertake work reaco st h leveBUGEd . St-LAURENan 2 l 2 reactorYA 1 d an s1 TA are in the stage of "definitive stop phase". That means all nuclear fuels have been discharged and transfered to the reprocessing plant. A stud s actuali y y runnin a possibl r gfo e decision concernin n immediata g e dismantlement at level 3 for MARCOULE G2 and G3 reactors. The question is economica technicad an l lF wito (whagraphite d EO hth o t e thow)d th ean r .Fo reactors, this level 3 is scheduled about 50 years after level 2 dismantling. Special attention is given to structure activation studies and evaluation of dose rates at a time when the technical support teams, in particular the neutronics teams, and the operating teams are still at the site or operational. It is effectively of first importance to prepare carefully files required fo dismantling which will occur after a such long time.

Other studie e beinar s g conducte o develot d p method r dismantlinfo s g problem - raising structures such as those comprising irradiated steels and graphites furnacc ar n t MarcoulA .e a e mad industrian ea l demonstratio meltinn no g more than 5000 tons of slightly contaminated steel from C02 primary circuit. The

51 benefit of this melting is evident concerning the volume of wastes, but also on the contamination level. Only cobalt subsists on the ingots which are use to manufacture waste containers. It have a capacity of 15 tons with possible production of 10,000 tones per year, for size pieces such as diameter 2.2 m, height 1.3 m. This furnace is actualy use r G2-Gdfo 3 steam generator melting fluidizeA . d graphite incineration pilot buil y FRAMATOMb t capacita s 0 kg/h4 Eha f o y, wherea e dermsth o laser system, installe t Marseillda collaboration ei n wit CEAe hth currentls i , y capablf eo incinerating 35 kg of graphite per hour.

2. HIGH TEMPERATURE REACTORS

Franc bees eha n monitorin developmene gth f Higo t h Temperature Reactors (HTR) throughout the world since 1979, when it halted its own HTR research and development program. The trend toward low power reactors with specific properties enabling the mo easilt y satisfy tighter safety requirement o studiet d le s s ha s assessing this concept. These projects fall unde scope wideth ra f eo r progran mo future reactors. The results of french studies are in good agrement with US and German studies on modular reactors for normal and accidental conditions.

France's interes demonstrates i t participatios it y db thren i y b e p CRPu t sse IAEA:

- 'Validatio f safeto n y related physics calculation enrichew lo n i s d gas-cooled reactors" in which a french expert was involved for the Proteus experimentations.

- "Heat transpor afted an tr heat remouva r gas-coolefo l d reactors under accident s involvei conditions" A benchmara CE n di e Th . k whic coda s hi experimeno et t comparaison between JAERI HTTR experimen frencd an t h calculation - sThes e calculations were made wit finitha e elements metho codD d3 e (TRIO-EF) whica hs i general code. With it, it is possible to have conduction, convection and radiation heat transfer! coupled. The results of calculation are in good agreement with HTTR experiment (scale 1/4).

"Experimenta- l wor validation ko f predictivno e method fissiofuer d sfo an l n product behaviour is gas-cooled reactors". In this area, the CEA undertook an experimental study in a research reactor (SILOE at Grenoble Centre) in the COMEDIE loop on behalDOEe deposith e f o .fTh fissiof o t n products generate particuley db s that were deliberatel tight studieds yno wa t , along with their migration during depressurization at various level performetess e wa tTh . d sucessfull lase th t n yquartei f 1992ro . Until today, the results were not published. As we have got DOE's agreement to release this information firsthie n o heath so wilt u w e subjectpieca rb lyo , ne f eo papeA . r signed by GA, ORNL and CEA will be presented during this IAEA meeting.

In addition, France examines the possibility of participation in a fourth CRP on "Design and Evaluation of Heat Utilisation systems for the HTTR". The final decision t takenye t i.sno

52 TABLE 1.

NET DATE DATE REASON REACTOR ELECTRICAL OF GRID CF SERVICE LIFE POR POWER CONNEXION SHUTDOWN SHUTDOWN

MARCOULEG2 40 22/4/59 2/2/80 21 Technica graphit: l e expansion

MARCOULEG3 40 4/4/60 28/6/84 24 Technical : steel embrittlement

CHINO1 NA 70 14/6/63 16/4/73 10 Economic

CHINONA2 210 24/2/65 14/6/85 20 Economic

CHINO3 NA 480 4/8/66 15/6/90 24 Economic

ST-LAUREN1 TA 480 14/3/69 18/4/90 21 Economic

ST-LAURENTA2 515 9/8/71 27/5/92 21 Economic

BUGEY1 540 15/4/72 27/5/94 22 Economic

VANDELLOS 1 480 6/5/72 19/10/89 17 Accident : alternator fire

TOTAL 2855 180

53 STATUS OF GT-MHR WITH EMPHASIS ON THE POWER CONVERSION SYSTEM

A.J. NEYLAN, F.A. SILADY '""""""XA9(642779~ General Atomics Diegon Sa , , California, USA

B.P. KOHLER AlliedSignal, Tempe, Arizona, USA

D. LOMBA GEC-Alsthom, Belfort, France

R ROSE Mechanical Technologies, Latham Yorkw Ne , , USA

Abstract The conceptual design of the Gas Turbine-Modular Helium Reactor (GT-MHR) has made significant progres e pasth t n i years . Evaluatio externan a f o n l versus internal (submerged) generato modificationd ran resulinternaa n a s a f s o t l seal task force were completed. Significant progres s alsswa o desig e generatoe madth th n f neo o r utilizing existing technology. Conceptual design of the turbocompressor was confirmed, including extensive evaluation of the entire turbomachine (turbocompressor and generator) rotor dynamics. Results concluded in a revised configuration for the location of magnetic bearings supporting the entire machine. Integration of the turbomachine with the recuperator, precooler, intercooler and internal ducts and seals progresse o improvt d e maintenanc d operationan e . This resulte n somi d e changed an s improvements in the overall arrangement of the power conversion module. The paper also provide summarsa safetfued e th an l f yyo assessment progress.

1. Introduction At the IAEA Technical Committee Meeting on Development Status of Modular High Temperature Reactors and Their Future Role held in November, 1994 at ECN, Petten, The Netherlands GT-MHR'e th , s design statu technicad an s l issues were reported t thaA . t time eth decision to focus the U.S. program on the direct cycle gas turbine power conversion had been made, the power level of the module had been selected, and design work was identifying the technical issues with the integrated power conversion module. AlliedSignal Aerospace (Tempe, Arizona bee,d USAha n )selecte turbomachine th s da e vendo presented an r papee dth r ) (Ref1 . discussin powee gth r conversion system design statu technicad san l issues. pase Inth t year AlliedSignal selected GEC-Alsthom (Belfort, France generato e woro )th t n ko r and Mechanical Technology Inc. (Latham, New York, USA) to work on the magnetic and auxiliary catcher bearings. General Atomics, the overall system designer, has integrated this turbomachine team wit e AlliedSignath h l (Torrance, California, USA) e heliueffortth n mo s recuperator and the ABB-CE (Windsor, Connecticut and Chattanooga, Tennessee, USA) work on the precooler, intercoole pressurd ran e vessel. Significant progres conceptuae th n so l desige th f no Power Conversion System bees (PCSha n ) accomplished. This paper update desige sth n GT-MHe statuth f so R with emphasi PCSe th . n so Section2 reporttrade th en so stud y that addresse e questiodth f whetheno generatoe th r r shoul e subdb - merged in helium within the power conversion vessel or be external to the vessel necessitating a

55 rotating shaft seal. The significant efforts by GEC-Alsthom to utilize their conventional hydrogen- cooled generator thir sfo s application als summarizede oar . Sectio addressen3 coupline sth e th f go generato turbocompressoe th o rt concomitane th d an r t change bearine th n si g locations latese Th . t analyses by MTI on the rotor dynamics are presented. Section 4 summarizes the findings of a task force comprised of PCS team members that addressed the demanding requirements for the sliding seals betwee componentS nPC limio st t helium leakage. Sectio capturen5 s improvement othen si r power conversion areas such as the precooler and intercooler configuration and the power conversion vessel layout. Sectio briefln6 y summarize statue sth othen si rGT-MH e areath f so R design development including an independent review by EPRI and its member utilities. Finally Sectio presentn7 conclusionse sth . 2. Design Progress on the Generator generatoe Th received ha r d relatively little attention prio GEC-Alsthoo t r m joinin teame gth . Cursory studie Generay sb l Electric Schenectad indicated yha d thainitiae th t l concept proposey db GA of a top-mounted, vertical generator submerged in helium, although non-conventional, was feasible (Ref . Earl2) . y concerns involve developmene dth t need largea f so , high speed generator immerse heliun di m withi pressurna e vessel versu morsa e conventional generator located outside the vessel. To address these concerns AlliedSignal took the lead on a trade study to investigate the advantage disadvantaged san externan a f so l generator wit rotatinha g shaft seal versu existine th s g submerged design. trade Th e study defined requirement rotatina r sfo g shaft seal betwee turbocompressoe nth r and the generator external pressure vessel. For this configuration the seal leakage must be held to stringent limits sinc leakage eth e consist f primaro s y coolant helium with potential radioactive contamination s assumewa t I . d that rotating shaft seal systemse werth t restricte n i eno t fi o dt confine existine th f o s g pressure vessel configuration, tha , thapressure is t th t e vessel coule db altered to accommodate any desired seal configuration. The trade study was initiated by requesting seal vendors to establish the feasibility of a low leakage helium retaining shaft seal. Both single shaft seals and buffered shaft seal systems were examined. The single seals included labyrinths rins facs ga g ga e y seals y sealsbufferee dr dr , d Th . an , d seal systems included combination f labyrintso t (oih s sealr waterwe seall o eithega r d s s sealso sy s an wa ga )dr r t I . found that a labyrinth seal by itself would not meet the leakage requirements. Dry gas seals, either alone or in combination with a labyrinth, would entail considerable development effort and risk, since existing designs hav mucea h smaller diameter than tha t mee te lifno requireth eto d d dan requirement combinatioseas A . ga l t witwe hlabyrinta a f no h woul r othed(o entail r oi rise f th lko liquid) getting into the primary system, and entail new helium-oil separation systems and possible mixed waste disposal needs. The study concluded that the problems associated with a rotating shaft seal are significant and would necessitate a major extension of existing technology to develop a suitable seal wit e requireth h d diameter leakagw lo , d lifeean , necessar e GT-MHth r fo y R application. At this point GEC-Alsthom cam boarn eo aggressiveld dan y addresse concerne dth s witha submerged design. They indicated that they do not expect significant problems operating vertically in heliu resultd man s from planned tests should confirm their expectations secone th s dA . largest generator manufacturer in the world they modified one of their standard hydrogen cooled generator operato st e verticall heliun yi m with magnetic generatobearingsy ke e Th r. requirements and their base providee sar Tabln di . e1 resultine Th g generato two-polea s i r , 360 synchronoum 0rp s uni MWet6 rate28 t da . This selectiomajoo tw rs nadvantagesha ) permit(1 : direca s t turbocompressor-to-generator coupling (i.e., freedom from the use of a gearbox or frequency converter), and (2) facilitates the use of the generator as a motor during plant startup. A brushless exciter system with a shaft-mounted exciter alternato shaft-mounted ran d diode uses s i supplyin r dfo controllind gan fielc d de gcurrenth e th i h t generator rotor. The overall generator conventional frame assembly (which can be removed and replace unita s da ) consist generatoe th f so r stato rotord an r , exciter stato rotord an r , magnetid can auxiliary catcher bearings, casing, and four water coolers. Heat is removed from the electrical equipment by means of conventional heat exchangers. Helium replaces the hydrogen coolant used in conventional generator excellenn thif sa o s i sizd ean t heat transfer media. Axial flow fane ar s mounte rotoe thesd th an rn deo circulate helium throughou e generatorth t . Tabl summarize2 e s

56 Table 1 Generator Key Requirements and Their Bases REQUIREMENT BASIS 1. Net power output (net power): 286 MWe Februar overaly94 l system point design. 2. Efficiency: Ove% r98 Energy balance at the February 94 point design. 3. Power factor: 0.85 Meets America internationad nan l market require-ment grir sfo d stability. 4. Output frequencyz H 0 6 : American market requirements. For 50 Hz markets, only minor changes are required in generator length and diameter, which can easile b y accommodate existinn di g vessel envelope. 5. Operate as motor during startup, Provide helium circulation without nee nuclear dfo r power. shutdown and reactor refueling 6. Vertical orientation Compatible witih turbocompressor orientation. 7. Magnetic bearings Eliminates possibilit l contaminationoi f yo . Compatible with turbocompressor bearings. 8. Submerged Eliminates need for rotating shaft seal. Helium provides for excellent heat transfer. 9. Two poles poleo Tw s allow higher speed (3600 rpm) operation than four poles (1800 rpm). Compatible with GT-MHR turbine speed. 10. 130% design overspeed Maintain structural integrit design yi n basic regime.

Table 2 Generator Design Features and Their Bases DESIGN FEATURE BASIS 1 . Synchronous Most efficient, most compact. 2. Solid, single forging rotor ("Turborotor") Turborotor is now the standard design in the specified power range. Turborotor design is stronger than salient pole design and generates lower windage losses. . 3 Brushless excitation Eliminates nee slir dfo p generatoe ringth n si r cavity. Slip ring causn sca e reliability problems. . 4 Asychronous exciter: Alternating curren uses ti n di Provides excitatio generatoe th r nfo r ove entire rth e the stationary exciter armature to induce alter-nating speed range, including startup zero speed condition, curren texcite e (ACth n )i r rotor thes i t n.I rectified without brushes. rotoe in th fee o generatore t dth r field winding. 5 . Armature voltage as a function of helium pressure Voltage levels are dictated by electro-magnetic design, in generator cavity compatible ar e with dielectric propertie heliumf so , For power operation, 19kV at >10 bar ancompatible dar e with GEC/Alsthom technology. For startup or shutdown, 4.2kV at 2 bar For refueling, IkV at 1 bar . 6 Heliu watemo t r coolers Conventional hydrogen design modifie heliumr dfo . 7 . Radial structural supports take stator reactions Conventional design. through flexible suspension 8. Axial structural supports mounted on Facilitates alignment. turbocompressor . 9 Startup drive statia y nb c frequency converteo t p ru This startup sequence is very close to GEC/Alsthom momene th t when turbine torque becomes positive practic largn eturbino s ega e applications. . 1Layou0 t includes conventional frame, cooling sys- Conventional design selections maximize the use of tem, electromagnetic circuit, windings, and flange proven technology. coupling

57 these and other design features of the selected generator and HEUtIM COOLED EXCITER their bases. Figure 1 provides OVERHUNG MOUNTED AC EXCITER a schematic of the generator MAGNETIC BEARING WITH CATCHER INDIRECT HELIUM COOLED STATUS MANHOLE within the top section of 4 AXIAL HELJUMAVATER COOLERS TERMINAL CONNECTIONS the power conversion vessel. RADIAL SUPPORTS Except for orientation and a GENERATOR CAVITr FLEXIBLE SUSPENSION few modificationn i e us r fo s ALTERNATIVE AXIAL-RADIAL DUCTS COKE-GRAIN aRENTED MAGNETIC SHEET heliu d witan mh magnetic )»CUUM DIRECT-COOLED HELD WINDING bearings, the GT-MHR gen- SHORT AXIAL CAHNNEt OPTIMIZE PACKETD DEN S erato s i almosr t identicao t l COPdS FLUX SHIM standara d GEC-Alsthom gen- AXIAL EXPANSION DEVICE AXIAL FAN WITH REMOVABLE BLADES erator seee . b Thi nn frosca m AXIAL SUPPORTS Figure 2 where the changes to MAGNETIC BEARING WITH CATCHER the horizontal hydrogen cooled MANHOLE ——————— generato r thifo r s application are highlighted. In summary, the generator configuration has Figure 1. Schematic of GT-MHR Generator received concentrated attentio confirno t initiae mth l design approach conventiona.A l generatos rha been employed to limit the development and testing needs to a few areas. By utilizing the traditional generator unit withi packagena d frame concerns regarding maintenance have been lessenede Th . vendor is supportive of the maintenance approach that has the generator removed from the vessel periodically (approximately every seven years when the turbocompressor is removed for maintenance accessed an ) visuar dfo hands-od an l n inspection. Startu operatind pan g conditions have been specified to limit the voltage at low helium pressure and dielectric strength. Tests are planned to confirm the behavior of the insulation at low helium pressures and during rapid depressurization from high pressure. A scale-up of existing power penetration technology to higher pressure will be required.

3. Design Progress on Turbocompressor As oppose e genth -o t d erator, design progress on the turbocompressor has not resulted in large visible changes to the layout. There has been no change in the arrangement of the turbine relative to the two compressors. Rather effort has focuse changen o d s introduced by adoptin a conventionag l generator and changes resulting fro techmy ke progres -o tw n so nical issues previously identified: rotor dynamics and differential thermal expansion. The initial turbomachine design whic s premisehwa n do Figur . Existine2 g Technology Applicabl GT-MHo et R very limited access to the generator cavity utilized a curvic

58 coupling between the generator and the turbocompressor. The coupling was located in the generator cavity whic separates hi d fro hottee mradioactivth d ran eturbocompressoe areth f ao a y rb labyrinth seal. A tie rod to the coupling extended from the top of the generator through the center of the generator rotor. Therefore, in this design access to the coupling was not required. However, r introduceba the eti d complex interactions with electronic exciter e leadth o t s , tensioningd an , vibration. With the decision to utilize a conventional generator, a conventional coupling was adopted requiring local access into the generator cavity. Providing adequate access was in itself highly desirabl consistend ean t with standard practic y GEC-Alsthomb e . Continuing wite th h approac alloo ht w accesgeneratoe th o st r cavity nexe th , t step involved relocatin axiae gth l thrust generatoe bearinth f o gp juso rt froto te mbeloth couplinge wth . Wit increasee hth d span between generatoe turbocompresoth e th d ran accesr couplinge rfo th o st secon,a d radial bearin addes gwa d so that there is one on each side of the coupling. Rotor dynamic differentiad san l thermal expansion were addressed firs observiny b t g thae th t area wit greatese hth t temperature difference s e colbetweet turbinth wa ho d s d e higan eth n h pressure compressor wher radiaea l firsbearine th t cooledf s tasko s ge wa wa s I On tackle. MT y db f acceptabli e se o t e rotor dynamics coul showe bearinge db th e generato f i th n n i s r cavity were change describes da d abovthif i sd turbine-compressoean r bearing coul removede db . Figur providee3 schematisa c comparin arrangemenw ne e coupline gth th f o bearingtd gan s with the previously reported configuration. In both cases the rotor dynamics challenge involves the relatively heavy generator rotor with an overhung exciter above and a long slender turbocompressor rotor below. The type of analysis that MTI conducted to determine the rotor bearing characteristics included: critical speed analysis, unbalance response analysi stabilityd san orden I . r assiso t determininn ti sourc e resonane gth th f eo t frequencies generator-turbine-rotoe th , r trais nwa analyzed for critical speeds and mode shapes by separating the rotating elements. This was accomplishe conductiny db independenn ga t critical speed analysi generatore th f so turbinee ,th e th , high pressure compressor pressurw lo e ,th e compressor combinee th d ,an d turbin higd ean h pressure compressor. Give mode nth e shapes fro individuae mth l components, worst case unbalance forces were applied in the generator, at the coupling, and at the low pressure compressor.

followine Th g conclusions were drawn fro rotoe mth r dynamics analysis. • The revised arrangement is a significant improvement ove prioe rth r concept. PREVIOUS NEW • The preferred approach to the bearing BOTH) EXCITER e highlus desigo yt s dampedi n w lo , stiffness bearings. This significantly THRUST BRG1 eases requirement supporr sfo t structures. BRS1 GENERATOR • Two resonant modes of the bearing-rotor GENERATOR system were encountered in relatively BRG 2 close proximit operatine th o yt g speed. COUPLING The modes are attributed to the generator • COUPLING THRUST

rotor at the exciter. > TURBINE e magnetiTh • c bearing neae turbinth r e TURBINE inlet hot zone was eliminated. COMPRESSOP H • R HP COMPRESSOR • Unbalance response analysis indicated that shaft amplitude e accuratelb n ca s y H H| 1 BRG4 BUG 4 controlled over the operating speed > LP COMPRESSOR COMPRESSOP I R range. • BRG 5 BRG5

Base thesn do e results considerable progress Figur . e3 Compariso Rotof no r Bearing has been made in understanding the turbomachine Schematics

59 rotor dynamics. Furthermore, programmable contro magnetie th f o l c bearing bees sha n utilizeo dt provide soft bearings with large damping that is well suited to the vertical rotor configuration. An improved configuration with less differential thermal expansion has utilized a conventional coupling. Future rotor dynamics work wil focusee possible b l th n o d e additio a magneti f o n c damper bearin overhune th t ga g exciter furthe.A r improvemen reduco t numbee eth bearingf ro s si the possible remova bearine th f lo g betwee compressorse nth .

. 4 Design Progres e Slidinth n o sg Seals Seals are extensively used within the PCS at component interfaces to maintain the integrity of the helium flow patassociated han d operating conditions. These conditions lea o significant d t pressure and temperature differentials across the seals, as well as to substantial differential motions of the components which each side of the seals contacts. The initial seal leakage estimates assumed a uniform leakage gap around the perimeter of each thoughts difficulwa e b p importano s t ga attaio sealwa e t t t Th n.bu plan o t t performance. Mucf ho the leakage was initially anticipated to be associated with the five horizontal turbocompressor seals and the vertical face seal at the turbine inlet. However, seal associated with the recuperator were found to be much more complex and without suitable provision for removal and replacement. The recuperator seal complexity resulted from the use of a single manifold to supply inlet flow to all six recuperator modules. This necessitated baffle plates above, below, and between modules to separate inlet flow fro outlee mth t systeflowa d an interconnectin;mf o g linear seals integrated with the baffle plates. The precooler and intercooler seals were also found to have significant complexities intercoolee .Th r seal complexity resulted from large diameter duct outlet seals which were not visible during initial installation and were not accessible for removal and replacement, as wel multipls la e water conduit leaf spring seals, omega seals bellowd an , s that wer t accessibleno e for removal and replacement.

sealA s task force comprisef do PCS team member s establisheswa d to address these problems wite th h objectiv f developino e g viable seal concepts that would resolv initiae eth l seal uncertainties e e scopth Th .f o e task force included those seals which were representativ mose th f te o diffi - cult seals to implement within the

PCS. The seals addressed by the task RECUPERATOR SEALS -= TURBINE EXHAUST force are shown in Figure 4. As a SEAL consequence of the activities of the task force which included input from seal vendors, major changes were made in the arrangements of the seals TURBINE INLET SEAL and oftentimes the associated INTERCOOUR WATER CONDUIT SEALS components. These changes are AND BELLOWS INTERCOOLER DUCT summarized in Table 3. OUTLET SEALS The recuperator was reconfig- ure includo dt dedicateea d manifold for each module tha s attachei t o t d the module during fabricationA . comparison of the recommended design with dedicated manifolds for Figur . eSeal4 s Addresse Tasy db k Force

60 Table3 Results of PCS Seals Task Force TASK FORCE RECOMMENDED LOCATION INITIAL CONCEPT CONCEPT ADVANTAGES RecuperatoP rL Common manifold for all One manifold per module. Eliminates linear seals and inlet modules. Baffle plates above, Single horizontal top plate baffle plates between below and between modules. above modules. Segmented modules Linear seals definin multiplga e cylindrical piston rings Uses same type seas a l sided polygonal cylinder. between module and top turbine exhaust p t I.Dplatea to d t .a an , plate. Turbine exhaust E-seal Segmented cylindrical Available in large diameters (representative piston ring. Accommodates relative horizontal al f o l f o motiont ou d san turbocompressor roundness seals). developmenw Lo t risk Turbine inlet C-face seal with pneumatic horizontao e Tw on d an l Eliminates face seal and actuated cylindricaf o l d piecen t ea vertical segmented cylindri- bellows hot duct. cal piston ring seals at ends Accommodates relative tee-shapew ne f o d adapter. motion Uses same type seal as turbine exhaust Intercooler duct Two large diameter inaccessible Single segmented cylindrical Reduced outer seal diameter outlet seals near duct inlet requiring piston ring seal at duct out- Inner seal eliminated blind installatio t shroudna s let interfacing with turbo- defining boundary between compressor. Seal added to Seal accessed through intercooler inlet and outlet, and turbocompressor seals, turbo-machine cavity between intercooler outled an t similar to exhaust seal and Uses same type of seal as precooler inlet. accessible through turbo- turbine exhaust machine cavity. Intercooler water Leaf seal at inner duct outlet Elimination of all seals All seals and bellows conduit outlet shroud. Welded omega seat a l through re-routin watef go r eliminated outer duct outlet shroud. Bellows piping. between outer shroud and tube sheet.

each recuperated module to the previous single common manifold design is shown in Figure 5. Based on analyses the dedicated manifold is not expected to impact recuperator effectiveness, however this will be confirmed in planned flow distribution tests. The initial baffle plates were eliminated and replaced with a single horizontal top plate. The initial network of linear seals was also eliminate replaced dan d with segmented piston rinplate p pistoe gto e sealTh . nth t rinsa g seals are the same type that are proposed for use at the turbine exhaust and turbine inlet. turbine Th e exhaus replaces seatE wa l d wit hsegmentea d cylindrical piston ring seal, which became typical otheal f ro l seals e turbinTh . e vertical inlet face s replaceseawa l d wito tw h horizontal segmented cylindrical piston ring seals betwee turbocompressoe nth adapterw ne a d . ran These seals, located above and below the hot duct, are the same type as the turbine exhaust seal. The tee-shaped adapter also interfaces to the hot duct with a segmented piston ring seal that replace duct ho t e bellowsth s previouslthas twa y use accommodato dt e thermal expansion. The intercooler duct outlet was designed to utilize a single seal at the end of the duct as a replacement for the two seals near the beginning of the duct. The new single seal interfaced with the turbocompressor, and therefore became another turbocompressor seal, which could be a segmented piston ring simila turbine th o t r e exhaus taccessibilite sealth t alsI d . oha y featuref o s

61 Dedicated Manifold 1 Per Module Figure 5. Compariso f Initiao n l Recuperator Manifold with Recommended Design Incorporating Six Dedicated Manifolds (One per Module)

turbine th e exhaust sealarrangemenn A . intercoolee th f o t alss owa r developed which eliminated watee th r conduit seal bellowd san thein si r entirety. With the above modifications, the complexity of many different types of unique seals was replaced with one type of accessible seal that has a proven history of use. This is a segmented cylindrical piston ring seal which incorporates both circumferentia r radial (o laxia d an ) l springo st maintain contact between the seal and its mating surfaces to attain a low leakage rate at low differential pressures, and utilizes the pressure differences to further reduce the leakage rate at higher differential pressures. This typ seaf eo shows i l availabls i Figurd n i an , ee6 fro t leasma t sourcesS U o tw : Stein Seal Cood san k Airtomic.

Wit e adoptio hth e sea th l f o n concepts proposed by the seals task force l seal al e accessibl, ar s r fo e remova replacemend an l t whee nth turbomachin s i removede e Th . seals will utiliz a esofte r material than those interfacing surfaces which are not intended to be replaced over their 60 year design life. The hardness of these latter surfaces wile b attainel d with suitable coatings, of which chromium-carbide is one candidate. A second sealing location on these surface s i incorporates e th n i d design as a backup should the first Figure 6. Typical Segmented Seal location become inadvertently

62 marred. This second surface woul usee db d only afte visuaa r l and/or optical examinatioe th f no first surface (via the turbomachine or vessel access penetrations) indicated such a need. Provision for removal and/or resurfacing of the "60 year" surfa.ce is also provided should the need arise. Although the task force changes have led to viable seal concepts which are expected to meet the design goals, outstanding issues remain. These will be resolved through further design and development. Testing will be conducted to establish and/or confirm the adequacy of seal and coating materials confiro t ; m seal design leakagd lif an eheliua n ei m environment under design conditions; and to confirm the effectiveness of the recuperator with the new seal and manifold configuration. The integration of the turbomachine seals with any flanges that are needed for turbomachine assembly wil resolvee lb d through further design effort actuae Th . l desige th f no seals wil performee b l proved dan confirmatorn ni y tests befor seale incorporatee th sar d intoa prototype PCS, where their performance will ultimately be demonstrated. The methods of detecting excessive inservice seal leakag beine ear g developed.

. 5 Design Progres f Otheo s r Power Conversion Components Effort bottoe th n si mpowee th hal f o fr conversion vessel were focuse volumn do e reduction and simplification decisioy ke A . n emanating fro seale mth s tas krouto t forc intercoolee s eth ewa r water outle intercoolete circuith f o p t froto r e inwarmth d radiall centee theth d o yt rnan dowe th o nt botto intercoolee th mf o r before exiting radially throug powee hth r conversion vessel y thin I .swa seals were eliminate intersectioe th t da heliue th f nmo wate e shrouth d r conduitdan previoue th n si s design tha intercoolere t th exite f o dp radiallto .e th Prio t y a thio t r s decision some thoughtd ha been give movino nt samprecoolee e th gth eo t elevatiop u intercoolere th s na side-by-sida ( r e concept) to save on vessel length and silo depth. While those benefits had some appeal, the eliminatio heliuf no m shroud-water conduit seal judges swa f greatedo r importance addition I . n sinc precoolee eth intercooled ran r heliu t significantla e mar y different pressurew lo e th s o t (due pressure compressor) supporte th f muc, o e har s different thicknesses which added complexity if both were in the same plane at the same elevation. Finally, maintaining the intercooler above precoolee th in-linn a n e(i r configuration t wit e watebu )th h r outlets routed back dowe th n center provides very similar designs for both finned tube helical coil water heat exchangers. TUReOMACXINE TUR80M6CHWE Vessel length and silo depth were saved by chang- e inpoweth g r conversion vessel's bottom head from a spherical n a desig o t n RKUP8WTOR RRUPERATOR elliptic one. This also allowed the precooler to be moved out radially which in turn decrease e preth d - cooler's height vessee Th . l and cooler improvements are best demonstrated in INTERCOOLER/ Figur compariny b e7 e gth PRECOOLER overall power conversion modules layout froe th m 1994 IAEAs a repor d an t integrated this year from September 1994 June 1995 each team members Figur . eCompariso7 Powef no r Conversion Modules

63 electronic files. Note also the overall increase in vessel height and the change in shape at the top improvee duth o et d generator definitio provisiond naccessan n ma r .sfo

6. Overall GT-MHR Design Status As described abov majoea r GT-MHe focuth f o s R desig developmend nan e lasth t n yeai t r was the PCS. However, progress was made in a number of other areas. Fuel plas formulateA .n wa utilizo t d provee eth n TRISO coating designs developed dan demonstrate Germand an bot n di S hU y e 198prioth o 7t r unsuccessful introductio f coatinno g desig procesd nan s changes pase th t n yeaI . r effor bees ha tn focuse improvinn do compactine gth g process relative to FSV experience so that particle SiC damage, both of a mechanical and of a chemical nature, is reduced. The SiC and contamination specification of 6x10-5 particle defect fraction in compacts was achieved by improvements in controlling particle forces during the compact matrix injection, reducin matrie gth x impuritie betted an s r conrol heae th t n i streatmen t process. The next capsule irradiation to confirm compact performance, MHR-1, was planned and designed. Demonstration compact e spesth ct tha werme t e manufacture e availablar d r an dfo e irradiation. The next step in the planned program is to modify the existing full size coater at GA and/or instal Germaa l n coater recently acquired from HOBEG. Kernels, manufacture t Babcocda k Wilcox& , woul coatee db d usin modifiee gth r acquiredd(o ) coater, formed into compacts using the improved process, and will be irradiated in a second capsule, MHR-2, to demonstrate the complete fuel system. Adopting this approac existine us o ht g technology sigificantly reduces risk and development costs. Spent Fuel Disposal. Options for GT-MHR spent fuel disposal were evaluated. Whole graphite fuel element disposal was preferred over separation of the fuel compacts from the graphite element because it involves less handling, the waste form is well suited for permanent disposal, angreatee dth r volume increases diversion resistance. Significantl s founwa t dyi thae wholth t e element disposal option does not adversely impact repository land requirements because they are determine decay db y heat loads. Transportatio whole th f ne o standara element e R us n dLW sca shippin coateC gSi caske d fueTh . l provides excellent long term retention s concludewa t I . d that GT-MHe th R wit TRISs hit O coated fue prismatin i l c graphite fuel element s weli s l suitea r dfo once-through fuel cycle wit additionao hn l processin f spengo t fuel. Plant Transien Safetd an t y Assessments startue Th . p strategy reported last yean i r confirmes wa Ref 3 . d with input fro generatoe mth r vendor generatoe Th . initialls i r y usea s da motor to provide mechanical shaft power for the turbocompressor. The generator cavity pressure is controlled to a high enough level for a given generator voltage to maintain an acceptable helium dielectric strength. The pressure is gradually increased so that a modestly sized static frequency converter can be used for motoring the generator. As the pressure increases, successively higher generator voltage levels are achievable until synchronization with the grid at 3600 rpm and 19 kV. More detailed steady state heat balances have resulted in power conversion component requirements that include margins for design evolution. Component designers have also received transient requirements for a number of planned and unplanned events, such as load following, reactor triplosd electrif san ,o c load. Accident evaluations specifi GT-MHe th o ct R confirmed thapassive th t e safety characteristics of the previous steam cycle modular high temperature gas-cooled reactor designs were maintained. Events initiated by one or more turbine blade failures were assessed. It was found that the resulting differential pressure forces across the prismatic core did not exceed the allowable graphite stresses. Since the dominant risk contributor for the steam cycle design were initiated by water ingress from e steath m generators e GT-MHth , s expecteRi havo t d a elowe r risk e publicprofilth o t e. References 4 and 5 provide more information on the GT-MHR safety evaluations. Utility/EPRI Review. A key event in the U.S. GT-MHR program was the review by utility/Electric Power Research Institute representatives held June 20 and 21, 1995 at General Atomics in San Diego. The review was prompted by an Advanced Reactor Corporation report that

64 characterize GT-MHe dth potentialls Ra "breakthrougya h technology s combinatioit o t e "du f no built-in safety and high plant efficiency. The ARC report proposed that a broader, intensive review conductee b utility db y representatives assemble EPRIy db . The objective of the review and subsequent report (Ref. 6) was to provide the reactor designers and other organizations with interest in future electric power generating options with an independent evaluatio e conceptuath f o n l GT-MHR design froe viewpoinmth f potentiao t l owner/operators e revieTh .w focusee PCS o paragraphth n Tw .o d s fro e summarmth e ar y reproduced below from their report. 'The review team concludes thae GT-MHth t R makes bolinnovativd an df o e us e emerging technologies that have the potential for much improved efficiency, simplification, and reduction of cost and waste. However, the current GT-MHR conceptual design needs substantial development, detailed design, experimental verification d licensinan , g effor e vieweb t n beforca d t i wite h confidenca s a e breakthrough technology that provides a technically and economically viable future optio potentiar nfo l utility customers." "The review team foun obviouo dn s show stoppers that would preven GT-MHe th t R from eventually meeting regulatory safety goal achievind an s mane gth y technicad an l economic goals set by DOE and the design team. Meeting these goals is viewed as both majoa r challeng prerequisita d ean GT-MHe th o et R joinin ALWe gth futurn Ri e U.S. nuclear power plant deployment." The utility design review provides very valuable input to the GT-MHR program. The review team identifie numbeda "challengesD needf ro R& r so " additiona thoso t l e identifie desige th y ndb team. The design team intends to carefully consider and respond to each recommendation. This utility input will provide the initial starting off point as the design progresses.

7. Conclusions e lasIth nt year considerable progres s beeha s ne GT-MH th mad n o e R programn A . experienced capable tea bees mha n assemble addreso dt desige sth developmen d nan powee th f o tr conversion system. Man desigy yke n issues have been addresse solutiond dan s identifiede Th . result f thiso s progress continu supporo et t earlier conclusions tha tfeasibilito thern e ear y issues and that continued desig developmend nan indeen tca d leabreakthrouga o dt h technology.

REFERENCES 1. Etzel, K., G. Baccaglini, A. Schwartz, S. Hillman, and D. Mathis, "GT-MHR Power Conversion System: Design Status and Technical Issues," GA-A21827, December 1994, Presente e IAEth t Ada Technical Committee Meetin n "Developmengo t Statu f Modulao s r High Temperature Reactor Theid san r Future Role," November 28-30, 1994, ECN, Petten, The Netherlands. . Cotzas . M . OrlandoJ 2 . McDonald. G R , d , "Heliuan F. , . C ,m Turbomachine Desigr fo n GT-MHR Power Plant," GA-A21720, July 1994, Presented at the ASME International Joint Power Generation Conference, October 8-13, 1994, Phoenix, AZ, USA. . 3 Rodriquez . ZgliczynskiJ , . C. Pfremmer, D d an , , "GT-MHR Operation d Control,an s " GA-A21894, Presented at the IAEA Technical Committee Meeting on "Development Status of Modular High Temperature Reactors and Their Future Role," November 28-30, 1994, ECN, Petten Netherlandse Th , . . . Dunn. 4 ShenoyNeylanD A . Silady . , A T J. , . . d "GT-MHF ,A ,an , R Design, Performance, and'Safety," GA-A21924, Presented at the IAEA Technical Committee Meeting on "Development Status of Modular High Temperature Reactors and Their Future Role," November 28-30, 1994, ECN, Petten, The Netherlands.

65 5. Dunn, T. D., L. J. Lommers and V. E. Tangirala, "Preliminary Safety Evaluation of the Gas Turbine-Modular Helium Reactor (GT-MHR)," GA-A21633, Presented at the International Topical Meeting on Advanced Reactor Safety, April 17-21, 1994, Pittsburgh, PA, USA. . 6 "Utility Industry Revie f Desigwo s nTurbine-Modula ProgresGa e th n o s r Helium Reactor," EPRI Report, September 1995.

66 __ i mil iiiiin mi inn HIM inn HIM inn \\ HTR PLUS MODERN TURBINE TECHNOLOGY XA9642780 FOR HIGHER EFFICIENCIES

H. BARNERT, K. KUGELER Research Centre Julich, Institut r Safetefo y Researcd han Reactor Technology, Julich, Germany

Abstract

The recent efficiency race for natural gas fired power plants with gas-plus steam-turbine-cycle is shortly reviewed questioe Th . competR n 'caHT e neth with high efficiencies? answereds i ' : Yes, it can - in principle. The gas-plus steam-turbine cycle, also called combi-cycle, is proposed to be taken into consideration here. A comparative study on the efficiency potential turbine-inles mades ga i C ° yieldt i ; 0 05 st 1 54.5temperature t a .% mechanisme .Th releasf so e versus temperature in the HTR are summarized from the safety report of the HTR MODUL. A short reference is made to the experiences from the HTR-Helium Turbine Project HHT, which was performed in the Federal Republic of Germany in 1968 to 1981.

Keywords: High Temperature Reactor HTR, modern turbine technology, gas-plus steam-turbine cycle, combi-cycle, efficiency natural gas fired power stations with gas-plus steam-turbine cycle, 3- pressure-steam-turbine cycle, release versus temperature, experiences from HTR-Helium- Turbine Project HHT.

HTR plus Modern Turbine Technology for Higher Efficiencies

. 1 Efficiency Race Triggere Naturay db s Ga l

1.1. In summary: The decline of the price of fossil energy carriers after the end of the oil price crisis, in particular the low price of natural gas, have triggered an impetuous development in gas turbine cycle technology. An efficiency race has been opened up to achieve higher values efficiencief o fossir fo s l fired power plantsparticulan i d ,an naturar rfo fires lga d power plants. The preferred solution of modern turbine technology is the gas-plus steam-turbine cycle technoloy, also called combi-cycle higA . h efficiency valu existinf eo g planta ; e.gs s% i 2 .5 typical value for the future perspective is 58 %. appetizen a s A 1.2 thir .rfo s chapte referenceso rtw :

1.2.1. Siemen , BereicsAG h Energieerzeugung (KWU): "The developmen turbines ga f o d t sdi achieve a new culmination at December 1994. During normal operation at our test site in Berlin the model V84.3A demonstrated performances which are number one in the world:

67 An , whicefficiencturbines % ga 8 ha 3 f f leadefficiencn )o y (o a o s t GUD-plan a f yo f o t 58 % (GUD = Gas- und Dampf-Turbine, Trademark) and

a power (of a gas turbine) of 170 MW, the largest in the class of the 60-Hz-turbines." Lit. SffiMENS-1995.

1.2.2. From a scientific report: "The efficiency can be increased from yesterday's promising 54 decadee th f o stepeac o n I .d tw valueh o en st n i stee th p% st threa aroun % e0 parameterd6 s are important, these are: higher effiencies by optimalization of the design of the blades, increased gas turbine-inlet temperature and improvements in the steam-turbine process, e.g. wit hsub-criticaa l 3-pressure-process including re-heat steps", lit. REEDLE-1994. 39 . S ,

detain I 1.3 efficiencien .o l turbines ga n o s s combi-cycle futurd san e prospect naturaf so l gas fired power plants:

1.3.1. In 1994 natural gas-fired power plants with gas turbines with the total power of about 240 GWe were in operation with a total efficiency of 32 %, and natural gas fired power plants with combi-cycles, tha gas-turbins i t e plus steam-turbine cycles^ST, wit totae hth l capacitf yo were operation eaboui GW 0 13 t n wit efficienc, RIEDLE-1994n ha % 9 4 f yo e , figTh . 1 . bigger number of capacity for gas turbines - in contrary to combi-cycles - is an indication that smaller unit capacitie smalled san r capital cost alse sar o decisiv decisioe th n a n e i r nfo investment. But there is a trend to make use of the potential for higher efficiencies with the combi-cycles, fig. .1

1.3.2. Natural gas fired power stations with steam cycles achieve efficiencies between 42 and 47 %, fig. 1. But obviously the gas turbine technology offers, in particular for natural gas based systems, a number of advantages, e.g. low capital cost, short construction time, and last not least high potentia efficiencyr fo l .

1.3.3. Official measurement powee th t sa r station AMBARLI, Turkey , plus witga sh a steam - turbine-cycle, GST, called GUD constructed ,an Siemeny db s resulte efficiencyn a n di 52,f o 5 % at nominal power, fig. 1, and 53,2 % at peak power, lit. SIEMENS-1993. This applies for firse th t block wit totae hth l MWepowe0 secon45 e f ro .thirTh d dan d block showed efficiencie nominat a s l powe 51.d 52.f . Thesran o 9 % 0% e measured values ,witt figfi e , hth .1 theoretical evaluations, lit. RUKES-199 REUTER-1993d 3an thed s an y, ga appl a r yfo turbine-inlet temperature of 1050 °C. The recently finished construction of the 205 MWe GUD power station TROMBAY, India, has a measured value of the efficiency at nominal operation of 50,48 %, lit. NAUEN-1995, table 1, at the air temperature of 35 °C, which adjusted to 24 °C means about 51,5 % at a gas turbine-inlet temperature (ISO 2314) of 1037 °C, lit. NAUEN-1995, tabl. e2

68 iJDAMBARLI Ut.: SlEHeMS-1933 TROHBAY 50 uv.

r Cb««bifo '

lit:. D focistirw Plant 15-- Project0 s f Sttwli«- 5 Stem T T; • zf» c

iittro- tooo _ •fortr i _* J/ loo looo . 1300 T °cr Fig. 1: Efficiency race natural gas, status and steae trenth r md fo turbin eturs cyclega -e th , bine cycle and the gas-plus-steam-turbine cycle: Net plant efficiency versus gas turbine inlet temperature, diagram 1 of 2.

1.3.4. With increasing gas turbine inlet temperature the efficiency increases at about 2 %- points/10 turbine-inles ga 0K t temperature thao s ,t aroun a t e b d n 120ca abouC 0% ° 5 5 t achieved turbines , figGa . .1 s wit turbine-inles hga 0 20 t temperature1 o t 0 10 range 1 th f n esi o introduction i w no e °Car n int markete oth perspectivee Th . thas si t aboua t yeae th t r 200e 0th gas turbine-inlet temperature could achieve 1 250 °C, lit. RIEDLE-1994.

1.3.5. It should be remarked here that the information about the promising value of efficiency of 58 % in the advertisement of the vendor industry - as usually - does not contain any precise informatio nturbine-inles abouga e th t t temperature; therefor respective eth e valu figs n ei i .1 tabled with a question mark. The reason is simple: The gas turbine inlet temperature is the simplest indicator for progress in the gas turbine technology.

1.3.6. Another factor to be considered is - of course - the capital investment.

69 witR TablhHT Gas-plu: e1 s Steam-Turbine Cycle Design2 , s 1 . Primary circuit Design A Design B Cooling fluid Helium Helium Thermal power MW 200 200 Reactor inlet temperature °C 350 364 Reactor outlet temperature °C 950 1050 Reactor outlet pressure bar 60 60 Relative pressure losses % 3,3 0,8 Helium mass flow kg/s 64,2 58,7 Electric blower power MW 3,3 0 turbins 2Ga . e circuit Working fluid Helium same Log. temperature differencX IH n ei °C 50 - Relative pressure losses % 6 3 Turbine inlet temperature °C 900 1050 Turbine outlet temperature °C 580 548 Relative cooling massflow % 3 3 Turbine mass flow kg/s 64,2 58,7 Turbine pressure ratio 2,42 3,71 Polytrope turbine efficiency % 90 92 Compressor inlet temperature °C 97 97 Compressor outlet temperature °C 291 396 Compressor mass flow kg/s 66,2 60,4 Compressor pressure ratio 2,58 3,91 Polytrope compressor efficiency % 90 91 Generator power MW 39 59,2 3. Steam turbine circuit Helium inlet temperature in waste heat boiler °C 580 548 Helium outlet temperature in waste heat boiler °c 97 97 HP - High pressure steam\HP °C/bar 550/60 515/140 LP -Low pressure steam\MD °C/bar 200/8 300/29,4 HD -Steam mass flow\LP kg/s 43,8 230/5,6 LP - Steam mass flow kg/s 6,9 - Polytrope efficienc steaf yo m turbine % 85 85 Generator power MW 57,9 52,3 4. Total plant Internal thermal efficiency of combined cycle % 48,9 55,7 Total power (sum minus mech. losses) MW 96,8 110 Total efficiency % 48,4 55,0 Self demand (with blower) MW 4,2 1,0 Net power MW 92,7 109 t efficiencNe y % 46,3 54,5

Remark: Desig : Lit.nA : HAVERMANN-1993, BARNERT-SINGH-1994 Desig : ThinB s study

70 2. HTR with Gas-plus Steam-Turbine, GST

2.1. In summary: The efficiency potential of the gas turbine technology for the conversion of high temperature heat from the HTR into electricity is - in principle - as high as that based on natural gas. Therefore it is proposed here to take the gas-plus-steam-turbine-cycle, GST, into consideration, in particular with a "3-pressure-steam-turbine-cycle". With further improvements particulaturbins n i , ga e eth cyclen ri witd assumptioe an ,h th s n ga tha e th t turbine-inlet temperature is 1 050 °C (100 K more than AYR in 1974) the calculated net efficienc particula54.s yi A . 5% rturbins versu T advantagga GS e se th cyclth f eo e with recuperation is that the core-inlet temperature is smaller at comparable efficiency conditions.

2.2. In detail on efficiencies from various projects and on the efficiency potential, in compariso conventionae th o nt l natura bases ga l d technology:

2.2.1 measuree Th . d efficienc f electricityo y productio THTR-30e th n i e ratie th th f oo s 0 a measured generator thermapowee th . rvs l powe 39,s ri (307% 3 MWe/763 nete MWtth -t no ,

Diagram 2 of 2 ?r cj. •>/-<- .^'(O WS St ' w// */< HRT-&ST

GfMHRO-n, / , lit-.ETZ6L-(M«- 4"../.t'iii.

Fig. 2: HTR with gas-plus steam-turbine-cycle, GST, trend efficiencn si comparisod yan n to natural gas: Net plant efficiency versus helium temperatur t core-outletea , respectively gas turbine-inlet, diagram 2 of 2: The HTR-GS same th s e Tefficiencha y potential naturae asth bases ga l d system.

71 efficiency), lit. SCHWARZ-1987 , tabl8 . , measuremenp ,e2 tpower mad% 0 e10 . These measured values were close to the calculated values. The THTR-300 had a steam turbine cycle cooliny andr da g toweroperates wa t core-outlei ,a t da e t temperaturTh , fig. 2 .°C 0 75 f eo operatio THTR-30e th f no terminates 0wa financiao t e 1989n d i politica d du ,an l l difficulties; rigth now (October 1995) the unloading of the fuel pebbles is finished since more than half a year.

2.2.2. The HHT-670 (HHT = HTR plus Helium Turbine, 670 MWe, projected demonstration turbine-inles ga a t a efficienc n planta % d 1 t 4 )ha temperatur f , lityo .°C ARNDT0 85 f eo - 1979 t shoulI . remarkede db , that this efficiency value coolingappliey dr projecr e sfo Th . t performes wa T dHH from Federa196e 198o th t 8n i 1 l Republi Germanyf co terminates wa t ;i d in 1981 because of the decision that the THTR-300 follow up-plant should be an HTR with a steam cycle. More detail givee sar litn i . WEISBRODT-1995, respectivel thin yi s workshop.

2.2.3. The efficiency of the GT-MHR (GT-MHR = Gas Turbine-Modular High Temperature Reactor) is 47 % at the gas turbine-inlet temperature of 850 °C, lit. ETZEL-1994, table 1. The turbins ga e cycle include compacsa t plat recuperatorn efi importann A . t design featur than ei t projecturbinheas e th ga td e thaexchangerss pari l tth ean f al t o t generatore wels th a , s a l e ar , located in the power conversion vessel.

2.2.4. The MHTGR-Combi (MHTGR = Modular High Temperature Gas-Cooled Reactor with Combine Turbins dGa Hypercriticae- l Steam Turbine Cycle, 238.n 5a MWe s ha ) efficiency of 53 % at the gas turbine-inlet temperature of 900 °C, fig. 2, lit. TELLIETTE-1993, fig. 11 on page 14, with an MHTGR of 450 MWt. It is foreseen to be operated with an intermediate heat exchanger, the core outlet temperture is 950 °C and the core inlet bastemperature thir Th efo s . proposa°C 2 MHTGR-HYPEn ea 51 s li R utilisin advancedn ga , hypercritical steam cycle alone with an efficiency of 48.4 % at the steam generator-inlet temperature of 730 °C, fig. 2. The interesting thermodynamical feature of this proposal is that the hypercritical steam turbine cycle allows a strong reduction in exergy losses in the steam generator: It fits very well to the heat cascading between the gas turbine and the steam turbine cycles.

2.2.5. Another possibilit reductioe th r yfo exergf no y losseheae th t n scascadini g betweee nth gas turbine cylce and the steam turbine cycle is the "2 or 3 pressure steam cycle" as used in modern conventional GST-cycles based on natural gas, chapter 1. Therefore it is proposed her tako egas-plut e eth s steam-turbine cycle, GST, into consideration particulan i , r - wit"3 ha pressure-steam-turbine-cycle".

2.2.6HTR-GSTe Th . turbine-inles ga efficiency n e ,a th desigs t a ha abouf , o % ntA 7 4 t thin temperaturi s, fig , C beinfigure° 2 °C . 4 0 2 g ,90 = adjuste frof Q eo T m r 46,d, fo tabl % 3 e 1, lit. BARNERT-SINGH-1994 and lit. HAVERMANN-1993. The HTR-GST, design A has the following main data: modular HTR-200 MWt-350 (core-outlet^inleC ° 095 , t temperature),

72 intermediate heat exchanger 900-291 °C, gas turbine with 2,42 turbine pressure ratio, steam turbine cycle wit pressureh2 internad san l thermal efficienc 48,f yo , tabl 9% colum, e1 n design A, lit. BARNERT-SINGH-199 litd .4 an HAVERMANN-1993 importann A . t design detaif o l design A (not given in table 1) is the pintch point temperature difference of 8 K. The higher values of efficiencies in fig. 2, about 49 % at 950 °C and about 50,5 % at 1 000 °C, were achieved by the omittance of the intermediate heat exchanger and the increase of the gas turbine inlet , littemperatur.°C HAVERMANN-19930 00 1 o et remarkee b . Hern ca t ei d that the tendency of these efficiency values does well fit with the tendency of measured efficiency value assistinf so g plants base naturan do l gas, called NG-GU figDn i . .2 witR detain hI HT 2.3 gas-plun .o l s steam turbine cycle, GST particulan i , r - wit"3 ha pressure steam turbine cycle", design B, this study:

2.3.1. The HTR-GST-design B has an efficiency of 54,5 % at the gas turbine-inlet , figtemperatur , thi°C .2 0 s study05 desig1 e f witeTh R .MWto 0 consistnhB 20 HT a n , a f so gas turbine cycle, comparable to the conventional ones and a 3-pressure-steam turbine cycle, taken from conventional design proposals figd , , figlitan .4 ..3 RUKESS-1993 , fig. 28 .6 . p , calculatee Th d design data, tabl , colume1 turbine-inles n ga desigmad e a ar r , efo n B t temperature of 1 050 °C, and include some improvements of the gas turbine technology as compared to design A. These are: polytrope turbine efficiency: 92 % (instead 90), polytrope compressor efficiency (instea% 1 9 : d 90), relative pressure losses (instea% 3 : , becausd6 e steam generator instead of compact recuperator) and relative cooling mass flow: 3 % (unchanged spitn i , increasef eo turbins dga e inlet temperature).

2.3.2. The gas-turbine-cycle of design B is similar to conventional ones, fig. 3. It has the following temperature data turbine-inles :ga core-inled t temperaturan , °C t0 temperatur05 e1 e 396 °C. It can be remarked, that the low core-inlet temperature is an advantage of the GST- cycle compared to the gas turbine cycle with recuperation.

2.3.3 3-pressure-steam-turbine-cycle Th . desigf e o bees , figha n , B 3 .take n unchanged from a conventional design, because it can be considered to be an optimum design already. The steam generator-inle steate temperaturth m, generato°C 8 54 s ei r outlet temperatur. °C 8 9 s ei The process also includes some re-heat. A possible disadvantage of the 3-pressure steam turbine cycle in comarison to the 2-pressure steam turbine cycle is the increased value of the 3rd pressure of 140 bar, which is higher than the primary helium pressure and therefore needs particularle b o t y considere design di n base accidents.

2.3.4 impression A . reductioe th n no exergf no y derivee losseb n sca d fro temperaturee mth - heat-diagram bundlee , steae figTh . th 4 .m n si generato arrangee rar sucn di wayha , thae th t temperature-heat-lines of the steam cycle fit best to that of the helium cycle. The pintch point temperature differenc desig n als s ei t bee ha ono nB change thereford (nodK an 2 1 t s ei

73 140bir/515*C .27,5b»r/515*C

0 10 0 5 0 ReCatn/e heat cascaded. Q% t/f} i for iK* 1 fr«s£ur« Skaw. Lit 8 euKes-tgg^ . f « C , i 3 for

Fig : Potentia3 . l HTR-GST wit efficiencn ha y Fig : Potentia4 . l HTR-GST wit efficiencn ha y of 54.5 %, flow sheet with gas-plus 3-pressure of 54.5 %, temperature-heat-diagram for the -steam-turbine-cycle, taken from conventional 3-pressure-steam-turbine for the flow sheet of designs turbins ga , e cycle adjuste reactor dfo r fig 3- the exergy losses of heat transfer are re- outlet temperaturC ° 0 05 1 f eo duced

mentioned in table 1) For future evaluations this value may be reduced to 8 K, resulting in a further improvement of the efficiency

even A n2. highe35 r efficienc HTR-GSturbine-inlef s yo ga a t a T% coul7 t 5 e db temperature of 1 150 °C, fig 2, dashed arrow and cross; the conditions is - of course - that a producn ca futur R e eHT hig h temperature 150hea1 f . o tC °C importann A 6 t23 advantag "gas-plue th f eo s steam-turbine cycle comparison "i "gae th so nt turbine cycle with recuperation thas core-inle"i e th t t temperatur smalles ei t comparablra e efficiency conditions. The following comparison can be taken as an example: The GT-MHR with 47 % at 850 °C, fig 2, has an core inlet temperature of 493 °C (919 °F), lit. ETZEL- 1994, fig , uppe.7 r part HTR-GSe corth ,n a s e , figinleha , T°C .2 t wit0 05 h1 54. t a 5% temperature of 396 °C, fig 3

2.3. addition I 7 thesno t e evaluation efficiencien so othed an s r desig ncoursf o data s i t ei , decisive to take capital investments into consideration.

74 238 The gas turbine-inlet temperature of 1 050 °C used in design B of this study is just 100 K higher than the mean core-outlet temperature of the AYR, which was achieved already in 1Q74

. 3 HTRs with Increase Outles dGa t TemperatureC ° 0 , e.g05 1 .

31 In summary The biggest challenge to increase the temperature of the produced heat is improvine th y b turbins R gga HT e technologth pun o t y Therefore future designR HT f so fuel on the basis of the TRISO-coated particle and of new HTR cores may realize increased gas outlet temperatures, e g 1 050 °C Reasons are good experiences of the operation of the AYR, other experimental results and proposals to increase the retention capability of the HTR fuel Another thae reasob ty contaminatio importann ma a e b t turbine no t th issuy f no es ma e a before

32 In detail on the retention capability of HTR-fuel in dependence from the temperature

321 In the THTR-300 the maximum fuel temperature of the fuel pebbles is 1 250 °C as the required design value and - at the same time - as a value for the commercial guarantee, fig 5, lit HKG-1969, Bd 1, S 4 16, tab 422-1 (THTR-300 Safety Report) For the release it is stated there "For these fuel elements the fraction of release for Xe-133 shall not exceed the meae 10'x valuth s n3 mea^ e a f ove e o th lifetim e s nra th ove d coree coatee ean rth Th " d particles of the fuel elements of the THTR-300 had a BISO-coating

gooA d 2 overvie32 retentioe th n wo n capabilit moderf yo n fuel pebbles with TRISO coated particles has been prepared for the HTR-MODULE describing the mechanisms of release of fission products, lit SIEMENS-INTERATOM-1988 (HTR-MODUL Safety Report), Bd 1

3221 In the lower range of temperature below about 1 200 °C the fraction of brakes of coated particles (expectation value alons )i e fabrication induced less i t si ,5 tha \Q~$,x g nfi 3 The hereby produced release is rather low and not very much dependent from temperature, lit S 3 2.2 1 -8 and 9

3222 At temperatures up to 1 300 °C no brakes of coated particles in material test reactors have been observed 321-S t li , 7

3223 With increasing temperaturesomC ° e0 diffusio30 1 o st abovn0 start20 e1 s from intact particles An additional fraction of brakes of coated particles in the range of 1 200 to doe C 160considereneet e ° b 0 sno o dt d accordin experimentse th o gt 3221-S t li , 9

322 temperature-inducee Th 4 d fractio brakef no coatef so d particle maximue th s a s- m value 8 1- 2 2 3 S t li , °C 0 60 1 10-x t 5 a 5s i -

75 t Liwi Hea i -1600 -\

1600° C: . 3.2.2.1-p * ^500-

diffusion

r -1300- >ho °c -1250 -1200<

(V" -

1000- = fract^bv f P ^ k<2ak.o C i ^ o s vaCu ^

m<*vT Cf> ^.ZAl-^-fig. -SZf.t/2.

gas ootutT

.-ttOl?Ut Safety Goo - U4.:

Fig. 5: HTR-MODUL fuel design data, release versus temperature, mechanisms of the release, and THTR-300 fuel data: From these data it is concluded, tha core-outlee th t t r temperaturfo C ° increasee 0 b 05 n 1 eca o dt future HTRs with moder turbins nga e technology.

3.2.2.5. The design limit for the heat up assumption is 1 600 °C.

3.2.2.6. The various ranges and limits of temperatures are illustrated in fig. 5 in linear scale.

3.2.4 temperature Th . e loa dModu R conditionHT e l areth r sfo : Mean gas-outlet temperature: 700 °C, mean power density 3 MWt/m^, lit. : S. 2.3.1-2, producing a maximum value of temperature of the coated particles of 830, respectively 837 °C, lit.: S. 3.2.4.1-4+ and fig.

76 3.2.4.-1/2. These value alse sar o illustrate conclusiona s figA n d i . .5 maximue :th m valur efo heat transfer in the fuel pebbles, as the difference between the maximum fuel temperature and the mean gas outlet temperature is at a power density of 3 MWt/m^ about 130 K, respectively 137 K.

conclusion I 3.3. futura r nfo e core outlet : temperatur°C 0 05 1 f eo

3.3.1 perspective Th . increasen a r efo d meae valuth f neo core-outlet temperaturR HT e th f eo with pebble bed core is the value of 1 050 °C. This follows from the information given in the above chapter 3.2. For this reason the evaluation of the HTR-GST-cycIe in chapter 2 has been don turbine-inles e witga e hth t temperature (being core-outlee equath o t l t temperaturef )o 1050°C.

3.3.2. Among the various applications of the HTR, (as e.g. steam cycle, steam cycle plus district heat or process steam, process heat applications for methane reforming and coal refinement othersd an , biggese )th t challeng improvo et temperature eth producee th f eo d heat improvine th y b turbins R gga HT ee technologyth n o t pu reasone s i relativele Th . th s si y strong influence on the efficiency, mainly because of its reducing influence on capital costs.

3.3.3. A possible contamination of the blades and other structures of the gas turbine due to the long term release of fission products, e.g. Cs-137, may in future not be an important issue befores a ,HHT-project e e.gth n .i , becaus technologiee eth remotr sfo e handlind gan inspectio wels n a othe s a l r maintenance conditions have improved.

. 4 Experiences fro Projece mth t "HTR with Helium-Turbine, HHT"

4.1. In summary: The project "HTR with Helium-Turbine, HHT" was carried out in the Federal Republic of Germany in 1968 to 1981. It has produced a large number of valuable experiences projece beeTh .d ha tn terminated becaus industriae th f eo l decisio projeco nt n a t HTR with steam turbine plant as the THTR-300-follow-up-plant.

4.2. In some detail on the HHT-project and the high temperature helium test plant HHV:

4.2.1 projece Th . t "HTR with Helium-Turbine, HHT", bein Federae g th carrie n i t l dou Republic of Germany in 1968 to 1981 in cooperation with the United States and Switzerland" and with the support from Swiss and German utilities had the objective to convert high temperature nuclear heat into electricity, using helium as the working fluid. Within that project to bigger test facilities were design and operated: The "helium turbine co-generation power plant, EVO" (EVO = Energie-Versorgung Oberhausen, Energy Supply at the City of Oberhausen in Germany) and the "High Temperature Helium Test Plant, HHV" (HHV = Hochtemperatur-Helium-Versuchsanlage, High Temperature Helium Experimental Plant, at Jiilich)A KF recentlA . y produced summar technicae th n yo l experience gives si lit.n i : WEISBRODT-1995, which is reported also in this workshop.

77 4.2.2. Within the HHT-project a project study has been performed for a demonstration plant HHT-670 with a net output of 670 MWe, a net plant efficiency of 41 % and with dry cooling. The gas turbine-inlet temperature was 850 °C at a pressure ratio of 2.84 and with a reactor pressure of 70 bar, lit.: ARNDT-1976. Details on the design are given in fig. 6 and on the cycle and turbo-machinery in fig. 7.

4.2.3 additionan A . l look, compare lito dt . WEISBRODT-1995 Hige th hn o ,Temperatur e Helium Test Plant HHV with respect to the turbo-machine, the test loop and the temperature- entropy-diagram is given in fig. 8, lit. WEISKOPF-1970. An interesting feature of the HHV- plan thas twa t heabroughs twa t int tese oth t loo compessope th onl a turboe yvi th f ro - machiner heaa a yt vi (anexchanger) t dno . Tha illustrates i t temperature th n di e entropy- diagra figmn . i 8 .

5. Summary and Results

5.1. Main result: Nuclear energy, as a source for high temperature heat, e.g. from the High Temperature Reactor principl,n i HTR - sams e th ha , e- e high efficiency potentia naturae th s a l l gas-based conversion process, both using modern turbine technology. The most modern gas turbine technology is the "gas-plus steam-turbine-cycle, GST", which also can be used in a closed form. A comparative study shows: At 1 050 °C gas turbine-inlet temperature the net efficiency is about 54 %.

Tr Lit: (*.. coetmj

Rg. 6. HTR Heliumturbine Proj. HHT 1968-1981 HeliumturbinR FigHT . 7 . e 1968-198T PiojHH . 1 Project-Studie Demo-Plan 1970 67 9 T HH t Flowshee Turbomachind an t e

78 r

Pr - 4-5 MW 0 M3 W = c ? H5 Wt- P - M K" *5 HW

. Hig8 . h Rg Temp Helium Test PlanV HH t

detailn I summar5.2a s . A : previouf yo s chapters:

5.2.1. The decline of the price of fossil energy carriers after the end of the oil price crisis, in particula priw naturaf clo o e rth l gas, have triggere impetuoun da s developmen turbins ga n i t e cycle technolgy efficiencn A . y rac bees eha n opene achievo t p du e higher effiencie fossir fo s l fired power plants, and in particular for natural gas fired power plants. The preferred solution moderf o n turbine technolog "gas-plue th s yi s steam-turbine cycle, GST" also called combi- cycle. A high efficiency value of an existing plant is e.g. 52 %; a typical value for a future perspectiv% 8 5 s ei

5.2.2. The efficiency potential of the gas turbine technology for the conversion of high temperature heat from the HTR into electricity is - in principle - as high as that based on natural gas. Therefor proposes i t ei d her tako e t "gas-plus-steam-turbine-cycle e eth , GST", into consideration, in particular with a "3-pressure steam turbine cycle". With further improvements, in particular in the gas turbine cycle, and with the assumption that the gas turbine-inlet temperature is 1 050 °C (100 K more than AYR in 1974) the calculated net particulaA efficienc% 5 4 r 5 advantag s yi GST-cycle th f eo comparison ei "gae th so nt turbine cycle with recuperation" is that the core-inlet temperature is smaller, at comparable efficiency conditions.

79 5.2.3. The biggest challenge to increase the temperature of the produced heat is put on the improvine th y b turbins R gga HT e technology. Therefore future e th designfue R n o l HT f so basis of the TRISO-coated particle and of new HTR-cores may realize increased gas outlet temperatures . Reason gooe °C ,th 0 e.gde 05 s experiencear .1 operatioe th AYRf e so th f no , other experimental results and proposals to increase the retention capability of the HTR fuel. Another reason may be that the contamination of the turbine may not be an important issue as before.

5.2.4projece Th . t "HTR with Helium-Turbine, HHT Federae th carries n i " wa t l Republidou c Germanf o 196n yi 1981o 8produces t ha t .I largda e numbe valuablf ro e experiencese Th . project had been terminated because of the industrial decision to project an HTR with steam turbine THTR-300-follow-up-plant e planth s ta .

BIBLIOGRAPHY

ARNDT-1979 Arndt, E., Biele, B., Kirch, N., Sarlos, G., Schlosser, J.: HHT-Demonstration Plant, European Nuclear Conferenc '79C ,eEN Hamburg, 6-11 May, 1979.

B ARNERT-SINGH-1994 Bamert , Singh,H. Futur: J. , e Application HTRf so : Electricity Productio Procesd nan s Heat Applications, IAEA Technical Committee Meeting "Development Status of Modular HTRs and their Future Role", November 2S.+29. 1994 and ENC Workshop on the Role of Modular High Temperature Reactor Netherlandse th n si , Novembe Decembed an 0 r3 , 19941 r , ENC, Energy Innovation, Petten, The Netherlands. ETZEL-1994 Etzel, K., Baccaglini, G., Schwartz, A., Hillman, S., Mathis, D.: GT-MHR Power Conversion System: Design Statu Technicad san l Issues, General Atomics, Allied Signal Aerospac- eGA A21827, Dezember 1994, IAEA Technical Committee Meeting "Development Status of Modular High Temperature Reactors and their Future Role", November 28-30, 1994 ECN, Petten, The Netherlands.

HAVERMANN-1993 Havermann , BarnertM. , , SinghH. , Thermodynamisch: J. , e Optimierung einer HTR-Kombi- Anlage, Forschungszentrum Julich GmbH, KFA, Institu Sicherheitsforschunr fu t d gun Reaktortechnik, Interner Bericht, KFA-ISR-IB-5/93, Marz 1993. in English: Thermodynamical optimizatio HTR-Combi-Plantn a f no , Research Centre Julich GmbH, KFA, Institut Safetr efo y Researc Reactod han r Technology, Internal Report, KFA- ISR-IB-5/93, March 1993. HKG-1969 Hochtemperatur Kernkraftwerk GmbH: 300 MWe THTR Prototyp-Kernkraftwerk, Sicherheitsbericht, Konsortium THTR, Brown, boveri & Cie AG, Brown Boveri/Krupp Reaktorbau GmbH , Aug, 2 Band d .un 1969 e1 ; In English: 300 MWe THTR Prototype-Power Plant, Safety Report, Consortia THTR, Volumes 1 and 2.

80 NAUEN-1995 KraftwerD GU W kM TrombayNauen5 20 : A. , : deutsch-indische Kooperation, BrennstoffTWarme-Kraft, BWK, Band 47, Nr. 7/8, S. 289-294, July/August 1995. in English: 205 MW GUD Power Plant Trombay: German-Indian Cooperation.

REUTER-1993 Reuter, F.D.: Dampferzeuger fur Kraftwerkskonzept l hoheemi m Wirkungsgra VDI-1993n di , S. 125-140; in English: Steam generator for concepts of power plants with high efficiency.

RJEDLE-1994 Riedle , VoigtlanderK. , , WittochowP. , : FrageE. , Nutzunr nzu g fossiler Energietrager de d run Klimarelevanz VDI-1995: in , . 35-54S , ; in English: Question to the application of fossil energy carriers and the relevance of the climate.

RUKES-1993 Rukes : KraftwerkskonzeptB. , fossilr efu e Brennstoffe VDI-1993: in , . 3-40S , ; in English: Concept r powesfo r station fossir sfo l fuels.

SCHWARZ-1987 Schwarz, D., Baumer, R.: THTR Operating Experience , 1st International Seminar on "Small and Medium-Sized Nuclear Reactors, Lausanne, August 24-26, 1987.

SIEMENS-1995 Siemen , BereicsAG h Energieerzeugung (KWU) 3A-Gasturbinen-Familie :Di Siemensn evo , Firmenschrift, 1995; in English Familie 3A-Gae Th : th f yo s Turbines from Siemens, Booklet fro Companye mth .

SIEMENS-1993 Siemen , BereicsAG h Energieerzeugung (KWU) GUD-Kraftwers :Da k Ambari Kraftwern ei , k setzt MaGstabe, Firmenschrift 1993; in English: The GUD-Power Plant Ambari marks the Future. Booklet from the Company. SffiMENS-INTERATOM-1988 Siemens/Interatom: Hochtemperaturreaktor-Modul-Kraftwerksanlage, Sicherheitsbericht, Bande 1-3, November 1988; in English: High Temperature Reactor-Modul Power Plant, Safety Report, Volumes 1-3, November 1988.

TILLffiTTE-1993 Tilliette, Z.P.: Modern Energy Conversion System and Nuclear Energy, Utilization - of a European, Commercial, Supercritical Steam Cycle, - of an Advanced Hypercritical Steam Cycle, - of Combined Gas-Steam Cycles, American Nuclear Society 1993 Winter Meeting, San Francisco, California, U.S.A., Nov. 14-19, 1993.

VDI-1995 Verein Deutscher Ingenieure, VDI, VDI-Gesellschaft Energietechnik: Kernenergie nach 2000, Tagung Aachen, 15.-16. Marz 1995, in English: Society of German Engineers, Society of Energy Technology: Nuclear Energy after the Year 2000.

81 VDI-1993 Verein Deutscher Ingenieure, VDI, VDI-Gesellschaft Energietechnik: Fortschrittliche Energiewandlun -anwendungd gun , Tagun . Marg 25 Bochum zd 1993un . 24 , ; in English: Societ f Germayo n Engineers, Societ Energf yo y Technology: Advanced Energy Conversion and Application.

WEISBRODT-1995 Weisbrodt. I.A.: Summary Repor Technican o t l Experiences from High Temperature Helium Turbo Machinery Testing in Germany, Prepared for IAEA under Contract 622 I 3-94-CT 1941, 94-CL9085, Reproduced by the IAEA, Vienna, Austria, 1995.

WEISKOPF-1970 Weiskopf: Hochtemperatur-Helium-Versuchsanlage, HHV, Firmendruckschrift Hochtemperaturreaktorbau GmbH, HRB, Mannheim, 1979; in English: High Temperature Helium-Test Plant HHV, Bookle Companye th f o t .

82 DESIGN OF HTGRs WITH CLOSED CYCLE GAS TURBINES

Session 2 DESIGN OF INDIRECT GAS TURBINE CYCLE FOR XA9642781 A MODULAR HIGH TEMPERATURE GAS COOLED REACTOR

Z. ZHANG, Z. HANG Institut f Nucleaeo r Energy Technology, Tsinghua University, Beijing, China

Abstract

This paper describe desigsa indirece th f s turbinno ga t e200MWe cyclth r efo t pebbld ebe MHTGR. In the design, the helium out of the Intermediate Heat eXchanger (MX) is extracte smala coolinV o dt RP l g s flowsystemga e s througTh . smalhrecuperatoa V RP l r and is cooled down, then it is used to cool the RPV. The whole primary circuit is integrated ipressurna e vessel core Th e. inlet/outlet temperature 550°C/900°Ce sar , whic suppln hca y s heaaga t sourc f 500°C/850°eo secondare th Cn i y sideheae Th .t source coul usee db o dt driv a nitrogee s turbinnga e plana cycl d tan e busbar electricity generation efficiencf o y about 48% is estimated. The thermodynamic calculation, preliminary design of the system components, and the important accident analysis are described in this paper.

Keywords: Reactor, Gas Turbine, HTGR

. Introductio1 n

The Modular High Temperature Gas Cooled Reactors can provide high temperature heat source such as 950 t with keeping the outstanding inherent safety. The electricity production efficiency of the MHTGRs by using Gas Turbine (GT) cycle can reach a remarkabl - 50%5 edirec4 e turbins .Th ga t e cycles, suc GT-MHs ha d an R T desigMI y nb GAI1]. Howeve e possiblth r e radioactivity depositio e turbinth n o ne blad d thue an eth s increas maintenancf eo e difficulties suggest that indirec turbins ga t e cycles shoul appliee db d firstly before enough experience is achieved to solve the radioactivity deposition problem in the direct cycle. majoe Th r difficult limio t yhighee th t r thermal efficienc indirece th f o y t lowee cyclth s ei r core inlet temperature. For the direct cycle, the cold gas leaving the precooler can be extracted to cool the reactor pressure vessel (RPV) and the other steel structures. Therefore the core inlet temperatur highes a e 500s b ra n -eca indirece 600°Cth r Fo t. cycles proposed before, suc MGR-GTs ha LidskyId proposean n core Ya ' th , ey d b inle t temperatur keps ei t 12 as lowe 310°s a rplane ordeCn Th i t busbacooo V rt RP l r efficienc MGR-GTf o y abous Ii t 42.1%. In order to achieve higher power plant efficiency, it seems special designs of RPV cooling shoul providee db indirece th r turbins fo dga t e cycle.

85 Auxiliary Blower

Nitrogen Reactor Vessel Cooler

500 C Nitroaen

Reactor Vessel Recuperator

850 C Nitroaen

Fig.l. Lavout of MHTGR-IGT

86 This paper provides a preliminary conceptual design of a indirect gas turbine cycle with a special RPV Cooling System (RPCS) The design, named MHTGR-IGT, has the following main features, e reactoTh r• Intermediatcoree th , e Heat eXchange CoolinrV (IHX)RP e g th , System (RVCS containee )ar reactoa n di r pressure vessel • High plant efficienc achieves i y d mainl y increasinyb g core inlet temperaturo et C ° 0 55 • Low RPV operation temperature is realized by a special design of RVCS

The thermodynamic analysis, preliminary design of the system components, and the important accident analysis are described in the following sections

. Syste2 m Design

As shown in Fig. 1, a 200MWt pebble-bed reactor core is located at the lower position of RPVe corTh e. geometr s samyi s tha a ef Siemen o t s 200MW HTR-Modular straighA . t tube IHX is located at the upper position of RPV and connected with the core through a gas duct. Similar to the AVR, the control rod system is installed at the RPV bottom and all controf o l rod insertee sar d upward intside oth e reflector maie Th .n helium blowen a d an r auxiliary blowe r shutdowfo r n topcoolinRVCV A . RP e locate Se gar th recuperato t da r (RVRRVCd an ) S cooler (RVC installee ar ) d respectivel annulan yi r regions outsidX eIH and blowers.

The cold helium from the main blower firstly flows downward through vertical channels in the side reflecto comed an r s int oplenua bottoe th t ma m reflector. Fro bottoe mth m plenum, the cold helium flows upward into the pebble-bed core and is heated up from SSO^c to 900 t, then the hot helium converges into a plenum at the top reflector and enters the IHX. In IHX, the hot helium flow upward through the IHX shell side and is cooled down by the secondary nitrogen t laste 550^A . th , c cold helium flows inte maith o n blower Brief technical data of the integrated MHTGR-IGT are given in Table 1. Detailed information on the design of IHX, RVCS will be given in the following section.

The GT cycle of MHTGR-IGT is similar to the MGR-GT design given by GA. In order to gain high plant efficiency, three stage compression "and two stage intercooling are used. The plant busbar efficiency of about 48% is estimated.

87 Table 1. Brief technical data of MHTGR-IGT Contents Parameters Reactor thermal power(MW) 200 Core power density(MWt/m3) 3.0 Core diameter(m) 3.0 Core average height(m) 9.43 Pressur primare th f eo y loop(MPa) 6.0 Primary coolant helium Core inlet temperature^) 550 Core outlet temperature(t) 900 Helium Flow Rate(Kg/s) 110 Working Fluid of GT cycle Nitrogen (N2) Pressur secondare th n ei y sid IHX(MPaf eo ) 6.0 Inle t(°c X temperaturIH ) n i 2 N f eo 500 Outlet temperature of N2 in IHX(t) 850 flo2 N w rate (Kg/s) 529 heighV ) RP (m t - 30 RPV diameter(m) 6.

MHTGR-IG e TablcyclT th G f . eo e2 T design

Contents Parameters Working fluid Nitrogen(N2) Flow rate of the working fluid(kg/s) 529 Turbine inlet temperatureCt) 850 Turbine outlet temperature(T:) 525 Precooler inlet temperature^) 102 Compressor inlet temperature(t) 30 Compressor outlet temperature(t) 77 IHX inlet temperature(T:) 500 Stage compressorf so s

88 Tabl Continuede2 Contents Parameters Stages of intercooling 2 Pressure ratio of the every stage of compression 1 5845 Polytropic efficiency of the turbine(%) 91 Polytropic efficiency of the compressor(%) 91 Plant thermal efficiency(%) 525 Plant busbar efficiency 48

. Componen3 t Design

3 1, Intermediate heat exchanger(IHX)

A straight tubular tube-and-shell heat exchanger is selected for the IHX In order to satisfy the compact demand, the heat transfer tubes with small diameter, close pitch, inner spire as well as outer low fin are chosen for the IHX tube bundle Preliminary designs of the IHX are given in Table 3

Table3 Design ParameterX IH f so

Contents Parameters Heat transfer capability 200. (MWt) Fluid in shell side Helium Flui tubn di e side Nitrogen Type of construction Straight tube Surface geometry Fin tube with inner spire

Tube diamete thicknesd ran s 14X 1. (mm) Tube pitch 18. (mm) Number of tubes 45625 Heat transfer area 14239 (m2) Height of tube bundle 71 (m) Inner/outer diamete tubf ro e bundle 1100/4390 (mm) Outer diameter of the IHX shell 4400 (mm) X IH Heigh e th f o t about 10 m

89 3 2, RPV Cooling System(RVCS)

Calculation by using THERMEX code indicates the maximum temperature of the RPV is smaller than 300X therf auxiliaro i : n s ei coolin V yRP g leakagsysteo n d man e flow froe mth core structure However, in order to avoid thermal shock of leakage flow with high temperature(- fro SSOtV reactoe mth RP o )t r graphite structures coolinV RP a g, system (RVCS) is designed

The flow diagram of RVCS is shown in Fig 2 The system consists of a RPV recuperator(RVR), a RPV cooler(RVC), orifices and valves As illustrated in Fig 1 and Fig 2, a bypass flow is extracted from the blower outlet and flows downward through the tube side of the RVR, which is installed in the annular region outside the IHX, and is cooled down from 550^ to 250^ Then the cold bypass flow enters an annular channel outside the RVR and flows upward into the shell side of a little cooler (RVC) and is cooled form 250 190to ^t \9Q°ce cor,d th an e betweevesses p V coli enters ga d dRP ga e an e l sth n th vesseo tw heatee l th fro y db m \9Q°c <0 220Co t 22 bypast last e °A c th , flows sga s upward through some vertical pipe comed shelsan R slRV intsidreturn d e blowee oth ean th o st r maie inleTh nt technica RVCle datth f aaiveo s Si tabln ni e4

————————— 7 ——————————————————

GAP BETWEEN RPV & CV RPV fflK ^ COOLER > \ ————— 1 V OPEN VALVE W -» «^

* I o j BLOWER 4

> f | ——— J ——— I 1 OPENVALV1E CORF. f "\ vl,i —— >———— — r— Kr-^ * ^v_x ——^ AUXLIARY > Q \ / VALVE BLOWER ORIFICE RCRPF.RATOR f\X| l^N < ——————— — — CLOSED

2 Flo g CoolinwFi V DiagraRP e gth Systef mo m

90 Table 4. Main technical data of the RVCS

Contents Parameters 1. Parameters of RVR Thermal capability 6.8W 6M Structure Spiral tube with tube jacket Flow rate in shell side or tube side 4.4 (kg/s) Inlet/outlet temperature in the side 550/250 (°c) Inlet/outlet temperatur sheln ei l side 220/520 (°c) Heat transfer area 267 (m2) Number of assembly 98 Outer diameter of assembly 114 (mm) Height of assembly 4.44 (m) Outer/inner diameter of the recuperator 4.92/4.6 (m) 2. Parameters of RVC Secondary coolant Nitrogen Structure Spiral tube with tube-jacket Heat transfer area 24. (m2) Number of assembly 33 Outer diamete f assemblro y 100. (mm) Assembly height ) 1.(m 5 Outer/inner diameter of the cooler 3.8/3.) (m 6

3.3. Shutdown Cooling System (SCS)

Accompany wit RVCSe hth shutdowa , n cooling system(SCS proposes i ) papee th n o di rt provide a simple and reliable decay heat removal during normal shutdown period. The system is shown in Fig 3. It consists of an auxiliary blower, a cooler as well as a recuperator. Because the cooler and recuperator are also the parts of the RPV cooling system(RVCS), SCS is simple and its equipments have multi-function. The decay heat removal under the accidental conditions is depends on the passive reactor cavity cooling system. The SCS propose thin di s paper therefor t safetno s ei y concerned.

In case of normal shutdown, the inlet valve of the main blower is closed and all of equipment in the secondary loop stop at the same time. The auxiliary blower of SCS then start drivo st e helium flowing throus coree hth , shelIHXR lRV , side tubR . RVCeRV e th ,

91 side, and finally back to the core. In RVC, the decay heat carried by the helium is transferre secondarC nitrogee RV th e o th dt y n n i side meany B . f adjustinso g flow ratr eo inlet temperatur nitrogene th f eo e e deca core th th , th y removee yeb b f heao n t ca t dou SCS.

GAP BETWEEN RPV & CV

CLOSED VALVE SHUTDOWN COOLER BLOWER

CORE —o^ AL70LAKY

VALVE RCEPERATOR ORIFICE HXr-

OPEN VALVE

. Flo3 g w Fi Diagra Shutdowe th f mo n Cooling System

4. Preliminary Accident Analysis

Preliminary accident analysi f MHTGR-IGo s s beeTha n usiny carrieb t g ou dTHERMI X code resulte analysie th Th . f so s indicates, under rating operation condition maximue th , m temperatur f fueo e l element s aboui s t 980

REFERENCES

1) M. Lidsky, et al. Modular gas-cooled reactor gas turbine power plant designs , the * second JAERI Symposiu HTGn mo R Technologie s, Toky o Japa n, Octobe r 21-23, 1992 ) 2 L.Yan. LidskyM . L , , MGR-GTI indirecn A : t close s turbinga d cyclR e poweMG e r plant for near-term development, MITNPI-TR-040, iMIT, Aug. 1991

92 3) Wang Dazhong, Zhang Zuoyi, Study of the gas turbine/steam turbine combined cycle o modulana r high temperatur cooles ega d reactor. Journa Tsinghuf o l a Universityl vo , 33, No.S3, 1993, pl-8 4) V.N.Grebennik, High-temperature intermediate heat exchanger for HTGR heat transfer to consumers, prepared for the first research coordination meeting for the coordinated research program on design and evaluation of heat utilization systems for HTTR, JAERI, 9-11, November 1994 ) 5 C.F.McDonald , Heaal t te , exchanger design consideration r HTGfo s R plantse th , American Society of Mechanical Engineering, July 27-30 1980.

lofll 93 CONCEPTUAL DESIGN OF HELIUM GAS XA9642782 TURBINE FOR MHTGR-GT

E MATSUO, M TSUTSUMI, K OGATA Nagasaki Research & Development Center

S NOMURA Nagasaki Shipyar Machinerd& y Works

Mitsubishi Heavy Industies Ltd Nagasaki, Japan

Abstract

Conceptual designs direct-cyclethe of helium turbine practicala gas for unit (450 MWt) andexperimentalan unit (1200kWt) MHTGRof were conducted resultsthe shownas and below were obtained. powerThe conversion vessel thisfor practical furtherunit can down- be outsidesizedan to diameter comparedas height7.4ma of 22m of and with conventionalthe design examples. Comparison of the conceptual designs of helium gas turbines using single- shaft type employing axial-flowthe compressor twin-shaftand type employing centrifugalthe compressor shows that the former provides advantages in terms of structure and control designs whereas the latter offers a higher efficiency. In order to determine which of them should be selected, a further study to investigate various aspects of safety features and startup characteristics will needed.be typesprovide two can Either the cycle a of to efficiency 46 of 48%. The third mode natural frequencies of the twin-shaft type's low-pressure rotational shaft and the single shaft type are below the designed rotational speed, but their vibrational controls are made available using the magnetic bearing system. Elevation of the natural frequency for the twin-shaft type would be possible by altering the arrangements of its shafting configuration. As compared with the earlier conceptual designs, the overall system configuration can be made simpler and more compact; five stages of turbines for the single-shaft type and seven stages of turbines for the twin-shaft type employing one shaft for the low-pressure compressor and the power turbine and; stages26 compressorsof axial-flowdie for type withsinglethe shaft system and five stages of compressors for the centrifugal type with the twin-shaft system. An overall system configuration of the flange joint method to preclude leakages from gaps between the elements was developed, using a plate-finned recuperator and intercooler, and a helically- coiled precooler feasibility its with fins,low and shown. is developmentA program leadto to the commercial MHTGR-GT plant consisting of three phases including the fundamental design of commercial unit, demonstration of the components technologies and design of the demonstration unit, and fabrication and construction of the demonstration unit was also planned.

. 1 Introduction As one of the energy sources to provide solutions to the prevailing environmental pollutions (global wanning, industry-related pollution dilemmo t energd e )th an n ai y supplye th , high-temperature heat utilization system using the Modular High-temperature Gas-cooled Reactor (MHTGR) and the Gas Turbine Generator System are now under research and development. The gas-cooled reactor used for this power generation system has originally evolved froresearce m th developmen d han t U.S.A.e workth \ n showini l sa t ^e saliens git t features of; CD inherent and passive safety characteristics (meltdown-proof), (D availability of high-temperature higheata h® , thermal efficienc extremel@ d radioactivan yw ylo e releases, all of which combine to show a new promising reactor for the next generation nuclear power plant. Particularly, fuel particles of approximately 1mm encapsulated in ceramics, etc.. are used

95 e reactoth s a r fue provido t l a econtainmen t functio f radioactivo n e releases withi e fueth nl particles perse and the gas-cooled reactor is provided with its inherent and passive safety characteristics to maintain the safety of the reactor because the heat generated from the fuel can alreleasee b l naturaterme n di th f so l heat release alone fro surface mmetallie th th f eo c pressure vessel. Mitsubishi Heavy Industries, Ltd. has been promoting the research and development of the MHTGR under cooperation with Japan Atomic Energy Research Institute and the other associated organizations over many years. Also compane th ,contracto e th HTTe s yi th f Ro r (High Temperature Test Reactor) under developmen a nationa s a t l projecs ha Japan i td an n undertaken its design and construction works. Furthermore, we have been engaged in research and development works of both types of MHTGRs using the steam cycle and the gas turbine cycle and the energy utilization system involved in heat utilization and electric power generation. Among these research and development activities, this paper presents the conceptual designs experimentan a f o l model unit (heat input 1200kWt utilita d yan ) unit (450 MWt) for the gas turbine generator system including the outline of our study results of the major components used fundamentae th , l characteristic f theso s e main component s welsa s a l the developmental issues. It should be noted that the fundamental concept and configuration of the MHTGR-GT have previously been developed principally by General Atomics.

. 2 Working Fluids 1 Selectio2. workine th f no g fluids Several candidate working fluids have been investigated as the working fluid for the MHTG theid Ran r physical propertie showe sar Fig.l^Tabld n ni presene an th e1 n I lt study, inert helium was selected from these candidate fluids because of its excellent heat transfer property.

Table 1 Physical Properties of Working Fluids

Density Specific Heat S.H. Molecular Atomic Molecular Gas Thermal Gas Constant Conductivity (kJ/kgKC O t a ) Ratio Equation Number Weight 3 (kJ/k) gK (W/mK) (kg/m ) Cp Cv Cp/Cv Helium He 1 4.0026 2.0772 0.1462 0.17850 5.19 3.116 1.66 Argon Ar 1 39.498 0.20813 0.0163 1.783771 0.522 0.312 1.66 Hydrogen Hz 2 2.0159 4.1244 0.1683 0.089885 14.188 10.06 1.409 Oxygen 02 2 31.9988 0.25983 0.0266 1 .42900 0.917 0.655 1.399

Nitrogen N2 2 28.0134 0.29680 0.0242 1.25046 1.041 0.743 1.400 Air — — 28.964 0.28706 0.0242 1.29304 1.006 0.718 1.402 Steam HzO 3 18.0153 0.46151 — — — — — Carbon Dioxide C02 3 44.01 0.18892 0.0146 1.97700 0.826 0.631 1.301

~~20 U) Xe-He 10 Mixture o IT 0 10 0 8 0 6 0 4 0 2 120 140 Gas Molecular Weight Fig 1 .Relativ e Heat Exchanger Surface Are . Variouvs a s Gases (By courtes f Airesearcho y )~~

96 2.2 Features of helium s compareA d wit e otheth h r working fluids liste Tabln i d , heliu1 e s physicamha l properties; (D a large thermal conductivity, (2) a large specific heat, (D a large gas constant, (4) a small molecular weight and (5) a small gas density. In accordance with such physical properties, helium has the features of its working fluid; (1) compact design of heat exchangers used heaD ( , t drop (output large smalb a t n ea l)ca temperature difference larga D ( e, velocits yi available at a small pressure ratio, © it leaks even through a micro pore and (B) it has a large volume per unit mass. In the open cycle gas turbine, its size must be increased when the working fluid has a small gas density whereas in the closed cycle gas turbine, the turbine size can be reduced by elevating the base pressure, utilizing the excellent features of helium. Comparison of helium anr relativdai theio et r feature f workino s s turbinga ge s fluidshoweth i r fo sFigs.2-n n i o 1t (refe2 2- Tablo t r together)e1 caspressure a f th e o n I . e ratio 2.8 theoreticae ,th l velocite th f yo heliu approximatels i s mga y 2.6 dimensionless 2it timed an r s thaai f o t s flow rat 0.16s ei 5 times as great. Since the output is proportional to the square of the theoretical velocity and to the produc floe th w f o trate outpue th , t fro heliue mth m turbin s 1.1i e 3 time e cassth e usinr gai under the same working conditions. Because helium has a small molecular weight leading to its leakage throug micro-poreha leakags it ,general n i , eis , assume greatee b o dt r than e thosth f eo other fluids, but as shown in Fig.2-2, its leak mass flow is approximately one-sevenths of the air leakage.

~~2500 IOU CO e, 160 2000 - / o 140 O / 120 7= 1500- / —— G-/f/(PA)H. 100 GVf/(PA)A,— —— 80r / - 1000 - 60 ./ o 1 o 500i 40~~

2.3 Features of the turbomachine using helium as its working fluid opee Inth n cycl s turbineega turbine th , e inlet temperatur elevaten a o s t bee t eha d nse point of 1350^ and the pressure ratio to approximately 30 to improve the thermal efficiency. When this fact is considered, the output from the open cycle gas turbine can be approximately 3.53 time outpue sth t fro closee mth d cycle heliu s turbinmga e unde same th r e base pressure condition. Meanwhile, the output from the closed cycle helium turbine can be raised to a level comparabl outpue th o et t fro opee mth n cycl turbins ega settiny eb base gth e pressur t morea e than about 3.53 times the atmospheric pressure. The base pressure (compressor inlet pressure) of the gas turbine for the MHTGR has t approximatela bee t nse (2.5a at 5 y2 s outpu MPa it d s approximateli tan ) y seven timee th s output fro opee mth n cycl s turbineega . Althoug specifie hth c hea heliuf o t ms approximateli y five times tha f airo t s pressur,it e rati s weli o l belo e leve wth f thi o l s difference th d an e volumetric changworkine th f o e g flui s significantli d y small leadin smalo gt l changes it f o s flow paths. Due to this fact, the change of the blading height can be small from the first to the last stage compressorf so turbinesd san .

97 . 3 Outlet Temperature Reactoe th f so r The MHTGR has a negative reaction property that its reactivity is reduced as the temperature increases internae th f shouls I .som ga o lt releasee e db e du accidenta t dou l failures, resultin pressura n gi en elevate a dro d pan d temperatur fuee th l f particleso e e reactivitth , y would drop, causin thermae gth l outpu reducee b o t t severao dt l percen desige th f no t level. e MHTGTh s beeha Rn designe o providt d e so-calleth e d inheren d passivan t e safety characteristics; the heat capacity generated during this failure can all be released out of the outer wall of the pressure vessel. The outlet temperature (turbine inlet temperature) of the MHTGR has been set at &5Q°C such that the maximum temperature of the fuel during an accident can alway maintainee b s d below 160Q°C whic allon hMHTGRca e wth thin I .s paper, studs yha been performe basie th thi f sn o do s temperature SSO'C. However mose th , t recent researches show a potential for an elevation of the maximum temperature due to improved fuel particles, re-adjustments of the bypass flows in the reactor and an improvement of the cooling method for the reactor vessel, etc.

4. Cycle Calculations and Component Efficiencies 1 Cycl4. e efficiency calculations calculatee Th d cycle efficiencie showe ar s Fig.3n ni . Whe recuperatena s usedi r e th , cycle efficiency increases at the lower side of the pressure ratio as the recuperater exchanging heat capacity syste e increaseth s ma d pressursan e los reduceds si sees A .n fro theoreticae mth l value (component efficiency 100%) of the recuperated cycle, this is because a raised pressure ratio would reduc e exhauseth s temperaturga t e resultin smallee th n gi r heat recovere th y yb recuperater. The efficiency of the recuperated/intercooled cycle is above the theoretical efficiency of the simple cycle in a pressure ratio range below 3.5. Because the efficiency of the

~~•*>.- 70 - \ ^-...^ Theoretical Recuperated x x "~~"-~....^ /Interceded Cycle~~

~~60 Theoretical Recuperated ^ x Cycle~~

~~Recuperated =• 50 /Interceded Cycle c 5 IU 40 73 O 30~~

~~20 Core Outlet Temp 850 O'C Turbine Inlet Press. 7.1 MPa Compressor Eff 38.0 % 10 Turbine Eff. 92.0 % Recuperater Eff 950 % Cooling Loss 1 0 % Mechanical Loss 2.5 % Pressure Drop 3.2 %~~

~~1.0 2.0 3.0 4.0 5.0 6.0 Compressor Pressure Ratio PH ) P/ (- L Fig 3 .Cycl e Efficienc f Direco y t Cycle~~

98 feasible heat exchanger could be estimated at 95% and the system pressure loss at around 3%, a pressur s selectewa e8 ratid2. f frooo m Fig.3. Assuming that efficiencie componente th f o s s are identical, the efficiency of the simple cycle is approximately 20% and that of the recuperated/intercooled cycl 46%s ei , respectivel pressure th t ya e rati 2.8f oo .

2 Effectcomponen4. e th f o s t efficiencie pressurd an s cycle e th los en o s efficiency e effectTh f efficiencieo s maie th nf o scomponent s (turbine, compressor, recuperater, intercooler s turbinega e systee ,th th etc. , f o m) pressure loscompressod san turbined ran inlet temperatur e overalth n eo l cycle efficienc showe ar y turbine Fig.4n ni th f I e. efficiencye th , compressor efficiency or the recuperater efficiency changes by 1%, the cycle efficiency also changes by approximately 0.5%. If the turbine inlet temperature or the pressure loss changes e cyclth , e befficienc1% y y , respectivelychange1% y b 0.3y r sb o % . Becaus pressure th e e loss shows the most serious influence on the cycle efficiency, pressure losses through pipings, etc., are needed to be minimized.

~~Recuperator Eff. Compressor Eff Turbine Eff. Core Outlet Temp Pressure Drop Compressor Inlet Temp~~

~~-150 15.0 Component Effectiveness~~

~~Definition of Effectiveness for Core Outlet Temp. (Ti -To)/To X100 (%) Suffi 0 xBase 1 Eac, h Case~~

~~Fig. 4 Effect of Components Effectiveness on Cycle Efficiency (Direct Recuperated / Interceded Cycle)~~

. 5 T^pe Selection 5.1 Direct cycle and indirect cycle direce Th t cycle use workine sth greactoe fluith n i ddirectlo rt y driv turbine eth e while the indirect cycle heats another working fluid through heat exchange utilizo t r e this heated fluid to drive the turbine. The direct cycle system can have a high cycle efficiency realized because its system can be simplified with a high turbine inlet temperature being made available. The indirect cycle system is supposed to show a higher safety reliability because it can minimize a potential for pollution of the turbomachine by radioactive substances due to use of another working fluid through the heat exchangers, and the freedom of its design work can be higher because the working fluid for the turbomachine can be freely chosen. Also, in utilization of the high-temperature, the direct cycle must take the type identical to the indirect cycle type. Both have advantages of their own and the view of the system advantages can vary, depending on what factor shoule son d place emphasi. son Currently it is shown that the fuel particle has a high reliability and that the plant safety e assureb n ca d even wit e direchth t cycle system sincd e potentiaan , eth l feasibilite th f o y indirect cycle system coul direc e hige dth b f hi t cycle syste mrealizeds i conceptuae th , l design direce oth f t cycle syste bees mha n conducte presenr ou n di t research.

99 2 Single-shaf5. t typtwin-shafd ean t type e twin-shafTh t syste s mcomposei o independentw f o d t shafts e poweth ; r turbine directly connecte generatoe e th turbine th o dt d drivo an rt s compressorse eth . Because this selection allows a free selection of the rotational speed of the drive turbine for the compressor, elevated efficiencies of the compressor and its drive turbine can be obtained. However, it require installatioe sth motoa f no drivo rt compressoe eth r during startu ther d problea pan s ei m of the difficult control of its overspeed for various assumed accidents. The MHTGR has a heliu compressore mth tand kan alsn startee sca ob utiliziny db high-pressure gth e helius mga store thin di s tank^single-shafe th n I . t system rotationae th , l speede turbine th th d f so an e compressor are needed to set to the same speed of the generator, which enlarges outside diameters of the turbine and the compressor, reducing the blade height which leads to difficultie e desigth n f i efficiencsno y improvements. Also, assuranc rotationae th f eo l shaft stability can be a major problem because the shaft length is extended with the natural frequency of the shafting system being low. The aforementioned problems have been examined by performing outline designs of two cases of heat outputs 450 MWt and 1200 kWt. In the case of the heat output 450 MWt, the efficiency differenc e compressoth n i e r betwee twin-shafe th n t e single-shafsysteth d man t system is 1 to 1.5% and the efficiency difference in the turbine is around 0.5% and the efficiency improvement and the compact design are possible, but the benefit of the twin-shaft system is small in terms of the shaft system vibration. heae case th I tnf th eoutpu o t 1200 kWt single-shafe ,th t system show poosa r efficiency because of a reduced blade height, and because the shaft becomes very small and long, it is determined that maintainin shafe gth t system stabilit difficulte b n yca .

6. Design Requirements and Specification The design requirements have been decided based on our study results of the cycle efficiency calculations, etc., as described earlier and on the study results in U.S.A. ^4e t a1^. The results of these design requirements are also applicable to the cases slightly deviating from the given conditions specificatione Th . respective th f so e components derived fro r conceptuamou l design are compared with another data^ ' f in Table 2. Both data are nearly identical except for the details of estimate and distributio4n5 o> f6 >losses 7 .

7. Design of the Turbomachine The fundamental technologies required for the design of the turbomachine have already been proven technologie aerospacf so industriad ean turbines turbomachine ga lth d san e b n eca designed using those technologies. However, the gas turbine for the MHTGR requires a design simultaneously provided with the lightweight design for aerospace application and the long-term endurance capability for utility service, and there has been no such design experience, which can be deemed a major developmental issue. The turbomachine should most adequately be designed and its periodical inspection interval should be established by examining and analyzing official permi authorizatiod an t n law varioun si s countries, regulations regarding the periodical inspection, etc., proven lives of the components for industrial and aerospace gas turbines factore th d limio st an , t their lives, etc.

1 Turbomachin7. e Type With e turbomachinregarth o t d e type, different e compressotypeth f o s d theian r r advantages or disadvantages are shown in addition to the differences between the single-shaft and twin-shaft type describes a s d earlier. Although earlier conceptual designs have dealt with the axial flow turbine and the axial flow compressor, the axial flow compressor has an extended shaft lengtlarga d ehan numbe s stagesit f o r , which increase s costsit s . Accordinglyr ou , conceptual desig alss nha o been case donth e n employineo centrifugae gth l compressor where the numbe f requireo r d manufacturin e stageth d san g e curtaile b cost n pipind sca e dan th o gt intercooler is more easily made available. Although the common practice of the twin-shaft type

~~100 Table 2 Specification of MHTGR-GT~~

System MHI-GT GT-MHR MHTGR-GT Thermal Power MWt 450.0 600.0 450.0 Coolant Pressure MPa 7.07 7.01 7.03 Flow Rate kg/s 244.0 328.0 243.0 Core Outlet Temp. "C 850.0 850.0 850.0 Inlet Temp. r 494.0 497.0 494.0 Gas Turbine Power MWe 208.0 301.0 228.0 Net Cycle Efficiency % 46.3 47.2 47.7 Compressor % 88.0 90.0 90.0 Turbine % 92.0 92.0 92.0 Recuperator % 95.0 95.0 95.0 Effectiveness Mechanical Loss*1 % 2.5 2.5 2.5 Cooling Loss"2 % 1.0 0.3 0.3 Generator Eff. % 98.0 97.5 97.5 Pressure Loss"3 % 3.2 4.4 3.5 Inlet Temp. •c 850.0 850.0 850.0 Turbine Outlet Temp. •c 515.8 517.8 514.1 Pressure Ratio — 2.66 2.64 2.67 Outlet Temp. •c 35.0 33.0 33.0 Compressor Inlet Temp. r 86.5 110.8 110.8 Pressure Ratio — 2.80 2.80 2.80 Heat Load MWt 516.0 658.0 484.0 Hot Inlet Temp. •c 515.8 517.8 514.1 Recuperator Hot Outlet Temp. •c 108.0 131.1 130.9 Cold Inlet Temp. •c 86.5 110.8 110.8 Cold Outlet Temp. •c 494.3 497.5 493.9 Heat Load MWt 92.0 167.0 124.0 Precooler Inlet Temp. r 108.0 131.1 130.9 Outlet Temp. r 35.0 33.0 33.0 No. — 2 1 1 Heat Load MWt 65.2 65.2 41.4 Intercooler Inlet Temp. •c 86.5 110.8 110.8 Outlet Temp. •c 35.0 33.0 33.0 Note *1 Ratio of House Load to Gas Turbine Power for GT-MHR Rati2 * f Generatoo r Windag s Turbine LosGa o et s Powe r GT-MHfo r R Tota3 * l Pressure Los Turbino st e Inlet Pressure

is to use a common shaft for the compressor and its drive turbine and use the power turbine to drivcommoa generatore f o pressureth w e lo us n e e shafth eth , r compresso e stagfo tth f eo d an r the power turbin bees eha n propose mako dt e availabl compressoe eth r startugeneratoe th y pb r and to preclude its overspeed, and the latter method has been studied in this paper. 7.2 Compressors The results of our conceptual designs of helium gas turbines employed axial flow and centrifugal compressors are compared with the conventional designs in Table 3 and in Fig.5. In our desigcompressorse th f no axiae th , l flow typbees eha n employe single-shafe th r dfo t type axiae th ld flocentrifugaan d wan ltwin-shaf e typeth r sfo casese th tl periphera e typeal th , n I . l spee s beedha n numbee th raise d f compressoo dan r r stages reduce s compareda d wite hth conventional designs single-shafe case th f eTh o . t typ s beeeha n divide H.Pe th o .dt grou d pan the L.P. group wit intercoolen ha r provided betwee groupso tw e nth , consistin stage3 1 f f go so axial flow compressors, respectively H.Pe th r L.P.e groufo ,th .e d groupcase pth an f th eo n I . twin-shaft type, the number of stages has been reduced to five stages, one-fifth of that for the axial flow type employiny b , centrifugae gth l typeoutside Th . e diamete impellee e th th f o rr rfo centrifugal compressor is as large as approximately 1.6m for the L.P. stage and the maximum peripheral spee higs a approximatels hs da i H.Pe ni/th 0 r .ys 50 stage fo therefored an , , stainless stee titaniur o l m alloimpellee th uses y i r dfo r material. Also twin-shafe th , t type use shafe sth t configuratio e L.P th shafe r .th stag fo tf no e penetrating throug e H.P th shafe r hth . fo stagt e such thaturbomachine th t generatostartee e b th n y db eca r which drive L.Pe sth . compressot ra the startup to raise the pressure in the system.

~~101 Table 3 Turbomachine Salient Features~~

Power Generation Unit System Nuclea s TurbinGa r e Heavy Duty Aerodenvative IndustriaT G l Gas Turbine Plant MHI-MHTGR-GT GR-MHR MS900 1(GEF ) LM6000 (GE) Power (MWe)208 286 226 42 Working Fluid Helium Helium Air Air Recuperated Recuperated Thermodynamic Cycle and Intercooled and Intercooled Simple Cycle Simple Cycle Turbine Inlet Temp. (*C) 850 850 1,288 1,243 Compressor Pressure Ratio 2.8 28 15 30 MasRatw Ro se (kg/sec) 244 320 613 123 Machine Orientation Vertical Vertical Horizontal Horizontal Shafting Type Twin Single Single Single Twin Overall Length (m) 7.9 8.9 13 14.5 4.5

Overal) l Diamete(m r 2.8 2.8 28 48 L 2-5 Overal WeighT G l t (kg) 65,000 70,000 82,000 300,000 5,600 Rotational Speed (rpm3,600/6,00) 0 3,600 3,600 3,000 10,225/3,000 Compressor Centrifugal Axial Numbe f Stageo r s 3HP+2LP 13LP+13HP 14LP + 19Hp 18 5L P14H+ P Max. Tip Diameter (mm) 1,762 1.580 1,683 2,515 1,372/737 Turbine Number of Stages 2HP+5LP 5 11 3 P 2H5L P+ DiameteMaxp Ti . r (mm) 1,905 2,500 1,778 3,251 889/1,321 Tip Speed (m/sec) 506/359 471 335 510 476/249 Blade Cooling Uncooled Uncooled Uncooled Cooled Cooled Bearing Type Active Magnetic Active Magnetic Oil Lubricated Oil Lubncated Number of Bearings (Thrust/Journal) 2/4 1/4 1/4 1/2 1/6

15 o 9 A N 30- O NT a A aO) CO o rj o> 10 ~ 55 (8 o 20 •r A 55 * in O HHT i O HHT A £ A GEC •S A GEC Q. DRAGON DRAGON E D ^ 5 D. A o 10 - • GA (MHTBR) • GA (MHTBR) O GA (MHR) •5 9 GA (MHR) 'o 6 A MHI (Centrifugal) A A MHI (Twin) 6 A MHI (Axial) MHI (Single) i i A , Z 0 n 1 2 3

~~Compression Ratio by Compressor Turbine Pressure RatiT JT o~~

~~Fig 5 .Compariso f Numbeo n f o r Fig. 6 Comparison of Number of Turbine Compressor Stages for Direct Cycle Stages for Direct Cycle~~

7.3 Turbines resultr conceptuae ou Th f so l desigturbine th f comparee no ear d wit conventionae hth l design Tabln si Fig.6n i numbee e3 Th . f turbino r e stage fivs i s e single-shafe stageth r sfo t typseved ean n e twin-shafstageth r fo s single-shafe tth typer Fo . t type increase th e , th f eo shaft lengt bees hha n avoided, aime high-loaa t da d design twin-shafe Th . t type which shows allowance for the load and the number of stages is found more advantageous over the single- shaft typ termn ei f performanceso outside Th . e diamete turbine th f singlee 2.5rs o ei th r mfo - shaft type and 1.9m for the twin-shaft type. In either of the types, the gas temperature at the movin lesr go blad s C whic° e0 inle85 belos hi s i theat-resistan e wth t temperatur blade th f eeo

102 material, eliminating the necessity of cooling. For the moving blade, a wide-chord blade to provid a reducee d numbe f bladeo r d improvan s s vibration-resistanit e t strengta d an h lightweight hollow blade to alleviate stresses to the disc and the connection between the disc and the blade has been employed. Also, because the blade tip clearance is increased when a back-up bearing is provided accompanied with the employment of the magnetic bearing, the moving blade is provided with shround and tip-fins to minimize leakage from the tip clearance. The disc has a large diameter and the heat-resistant temperature of the material suited to such a large disc is approximately 6QQ°C, which requires the disc to be cooled. The performance drop due to this cooling is estimated to be 0.5 to 1%.

. 8 Heat Exchangers Heat exchangers required for the gas turbine are a recuperator, a preheater and an intercooler, and their high efficiency performances and compact designs are demanded. Table 4 lists specifications of these heat exchangers and Fig.7 shows configurations of the heat exchanger elements.

~~Tabl e4 Heat Exchanger Salient Feature~~

Unit Recuperator Intercooler / Precooler MHI GT-MHR Industrial GT MHI GT-MHR Gas Turbine Power (MWe) 208 2860 6 o t 0 1 204 286 WorkingJTukl (High/Low) He/He He/He Air/Gas He/Water He/Water Unit Thermal Rating (MW) 515 630 0 3 1o 5t 65.2/92 131/164 Exchanger Type Plate-Fin Plate-Fin Plate-Fin Helical Coil Tubular Numbe f Moduleo r s 8 6 2 to 4 2/1 — Row Rate (kg/sec) 244 320 024 o t 5 4 244 320 Hot Gas Inlet Temp. (K) 516 510 519 to 575 87/108 112/131 Pressure Difference (MPa) 4.5 4.6 0.9 3.6/4.5 4/2.3 Effectiveness 0.95 0.95 0.84 to 0.89 0.83/0.84 0.93/0.95 Overall Pressure ) Los(% s 1.6 2.0 3.5 to 4 0.2/0.2 0.47/0.75 Typical Surface Density(m/m) 714 1,906 624 488/295 500 He: Heat Transfer Coeff. (W/mK) LP : 1,857 L 2,55: P 5 LP :115 780/58H, e:1 5 1,250 HP : 1,876 HP : 3,120 HP :425 Wa: 8600/7230 Water: 11, 000 Thermal Density (MW/m) 8.9 17 1 4.0/0.57 3.6 Typical Flux (W/cm) 1.5 2.5 0.1 0.82/0.2 0.75 Material 316StSt 316StSt StS9 40 t 1/2CM/2MO 1/2CM/2MO : Hea 1 • t Transfe Sids Totao et Ga t rl AreRecuperatoHo f ao r Heat Transfer Section Volume •2 : Typical Specification of First Stage Intercooler •3 : Low-Fin Tube

~~(a) Recuperator (b) Intercooler (c) Precooler (Plate-Fin) (Wavy-Fin Flat Tube) (Helical Coil)~~

~~Fig 7 .Hea t Exchanger Elements~~

103 1 Recuperato8. r recuperatore th r Fo plate-finnee th , d compact heat exchange bees rha n selected becausea high heat transfer demandes i efficienclarg a % o t e 95 e heaf ddu o y t transfe f approximatelo r y . DependinMW desige 0 th 52 n gno specification, variou typen sfi selectee sar r plate-finnedfo d heat exchangers and they are broadly used as small heat exchangers. For this system, the offset fin whic providn hca e higher heat transfer efficienc s beeha yn employeda s e offsei n Th .fi t small typ f 1.9m eo heightn i m , O.lm platmn i e thicknes 1.0md fin-to-fin si an m n spacingt bu , this would pose no practical problems because the working fluid is high-purity, inert helium, which will eliminate concerns about oxidatio foulingr no .

8.2 Intercooler The intercooler is used to reduce power consumption for the compressors and improve the efficiency of the regenerative gas turbine system. In this section, helium in the pressure rise process of the compressors is cooled by the cooling water running through a wavy-finned flat tube. The system design has been done on the basis of installation of two intercoolers, and Tabl showe5 specificatioe sth firse th t f intercoolere stagno th f eo .

8.3 Precooler e precoolerth r Fo e helicath , l coil type heat exchanger whic s provehha n resultr fo s various machine type has been selected and the low-fin tube has been employed to reduce the height. If a plate-finned or fin-and-tube type heat exchanger is employed for the precooler, it can reduce the installation space for the precooler. This subject will be dealt with in the next researcr stagou f eo h work.

. 9 Stud Shaftinn yo Bearingd gan s Fig.8 show arrangemente sth f shaftinbearine single-shafse o th th d r ggan fo t typd ean the applied to the practical unit of 450 MW. There are several examples of the shafting arrangements of the twin-shaft type for the gas turbine, and the fundamental arrangement uses one shaft for the gas generator and the other for the power turbine, which can reduce the overall shafting lengt allod han rotationae wth l compressoe speeth f do re freel shafb o t yt selected, providing optio highea r nfo r efficiency thit bu s, arrangement requires additio motora f no , etc., to drive the gas generator at startup. As shown in the figure, the twin-shaft type in this design use e low-pressurth s e compressor shaft inserted throug e high-pressurhth e shaft becausa e common shaft has been designed for the low-pressure compressor and the power turbine for the compressor to be driven by the generator.

1 Shaftin9. g Fig.9 shows the calculated results of the rotational shaft vibrations. The first mode is the parallel mode, the second X mode and the third the bending mode. As shown in Fig. 9, the designed rotational speed L.Pe th . f so shafting single-shafe th f so twin-shafe t th typd ean t type are in a region of rotational speeds higher than the third mode and it is difficult to exceed the natural frequenc e thirth d f o ymod e whe l lubricateoi n d bearing e natura th usede t ar s bu l, frequency of the third mode can be exceeded if the magnetic bearing system which can control the shaft vibratio s employedni . However s desirabli t i , maintaio et designee nth d rotational speed belo e naturawth l frequenc e thirth df o ymode henced s an neede,i t i , deviso t d e th e elevatio e naturath f o nl frequenc y reducinb y e shafth g t e lengthe turbinth th d f o an se compressor in order to raise rigidity of the shaft bending. For the two-shaft type, the L.P. shaft is placed throug holloe hth w H.P. shaft becaus startue eth p usin generatoe gth s beerha n made available as already described. This causes the length of the L.P. shaft to be extended, resulting in a reduced natural frequency. If a common shaft were used for the gas generator turbine and the compressor wit e poweth h r turbine separated from this commo e highth nf i - shafd an t pressure helium storeheliue th n dmi tank wer startupe th use r lengtL.Pe e dfo th th ,. f ho shaf t woul e reducedb d , causin e naturagth l frequenc e thirth df o ymod exceeo t e e designeth d d rotational speed. In this paper, the most serious problem of shafting has been examined.

~~104 10"~~

~~Rotational Speed 360m 0rp 103 5 10 10" Support Rigidity (kgf/cm) a) Single-Shaft Type~~

~~10" •o <]0>} Q. CO- 103 Rotational Speed 1st 2o 3600 rpm 10? 103 10" 105 106 Support Rigidity (kgf/cm) P (bShafL )f Twin-Shafo t t Type~~

~~104~~

~~wQ. 103 1st Rotational Speed 600m 0rp 10! U103 10* 10s 106 Support Rigidity (kgf/cm) ShafP (c f H Twin-Shaf)o t t Type~~

~~(a) Single-Shaft, Axial (b) Twin-Shaft, Centrifugal 9 Calculate g Fi d Resultf o s Compressor Compressor Rotational Shaft Vibration 8 Heliu g Fi m Turbine Rotor~~

9.2 Magnetic bearing magnetie Th c bearin provides gi d wit followine hth g features lubricano n ® ; neededs i t , © a small loss, (3) the shaft vibration is controllable and (J) the load capacity does not depend rotationae th n o lclosee speedth r d bearins Fo cycl. ga turbines e ega th g ) usin(£ ,workin e gth g fluid as a lubricant and (?) the non-lubricant magnetic bearing can be pointed out as candidate bearing avoio st d ingres lubricane th workin e f so th o t s bearing ga fluide gTh .pose s problems cooline oth f g metho smald dan l deformation e shafbearine th th tf d o sbecausgan s ga e eth bearing has a small load capacity and a large loss. Meanwhile, the thrust bearing of the magnetic bearing syste supporn mca larga t e load becaussmala s ha l t losprovided ei an s a s large suppor journate areath n I .l bearing, contro shafe th f t lo vibration higo t hw s lo fro e mth mode s availablesi . Therefore s besi t i , t bearin e suiteth r dfo g verticasystee th r mfo l shafting arrangement of an extended shaft length where the thrust bearing supports an overall weight of the rotor. The selection chart of the outside diameters of the magnetic thrust bearing is shown in Fig.10. Althoug dimensione hth journa e thrusth e th d f lan tso bearing ovee range ar s th r f eo proven experience, the stiln manufacturee yb ca l termn i d f technicao s l capability. However, the magnetic bearing is supposedly needed to be provided with ball bearings as the backup

~~105 10 Maximum N = 3600 rpm Speed~~

~~o u. 01 c £~~

~~100 150 200 250 300 Rotational Speed U (m/s) 0 FigLiftin1 . g Forc f Magnetio e c Thrust Bearing Relative to Peripheral Speed~~

bearings at the shaft ends in order to avoid contacts at startup and stop and a large clearance must be provided to avoid the contact of this bearing with the shaft during the normal operation. Accordingly t becomei , s necessar provido yt e large clearance turbinr sfo compressod ean r tips, which reduces their performances a s backudesirabli e t I us .p o t epowe r suppln a s a y alternativ backue th o et p bearin desigd gan n alleviatio f damagn o contac e th o et t area between th bearine e th shaf theid d gan an t r development expectee sar d for.

10. Design of the Overall System Configuration The structural sections of the utility unit and the experimental model unit as designed in utilitthie th s r researcy unitFo . configuratioshowa e , h12 ar d Figsn i an generato e 1 .th 1 f no r s turbinga ane deth containe modulee on n di recuperatoe th , modulee on n i rintercoolee th , n i r one preheatemodule th module e d on ean n s beei r ha , n employe maintenance th carro d t t you e servicing and the periodical inspection after removing each module from the pressure vessel in order to minimize the required work in the vessel. Flanges will be provided between the respective modules to preclude gas leakage from between the modules. The flange connection method must take into account thermal expansions and deformations, etc., of the assemblies and components thed yan , wil examinee lb detai e stage th th f len do i desig n work. supporte Th turbins structurga e eth r wil retainee fo b l suppora n do t plate betweee nth generator and the turbine and will use the hanger type. Accordingly, the casing structure of the gas turbine is to have a rigidity required to support the shafting and to be of the self- containment structure which could contain fragments of a damaged blade if such a damage should happen. assemble th r Fo y metho modulese th f do flange th , e connectio insere th r ne to b typ n eca conceived forme e problea alleviatios Th e . th ha r mf o n metho stressef do deformationo t e sdu s cause thermay db l elongation, etc., whil problea latte e sealine s th eth ha r f mo g methodn I . addition arrangemente th , componentf pipinge so th d san s mus carefulle b t y designed, taking into considerations acces f maintenancso e personne vesselse th o workine t l th , ge spacth d ean workability, etc.

11. Development Program The main flow of this development program is shown in following table. The preliminary research work in the initial period in this program has already been shown in this e overalth pape d lan r system configuratio outlinee th d an nd configuration e respectivth f o s e components have conceptually designed ,existin e baseth n do g technological informatiod an n MHI' design sow n database nexe Phasn th I t. n stepi fundamentae 1 th , l desigutilite th f yno unit will be performed and the component technologies and the design technology required for the development of the utility unit will be fulfilled. In Phase 2, verifications of performances

~~106 Generator~~

~~Generator Turbine LP Turbine HP Turbine Power \ HP Turbine Compressor~~

~~HP Compressor Intercooler 8 Gene. Turbine~~

~~Compressor Intercooler LP Compressor Compressor~~

~~Intercooler Recuperator~~

~~Recuperator "~ Recuperator~~

~~Precooler Precooler Precooler Starting Motor~~

~~(a) Single-Shaft Type (b) Twin-Shaft Type *1400~~

Fig. 11 Power Conversion Modules 2 1201 Fig t .Tes 0kW t Model Unit and functions of the required components as well as the design of a demonstration unit will be carried out, read respono t yexaminatioe th o dt e Safet th f yo n Counsil, etcPhasn I . , 3 e fabrication and construction of a demonstration plant will take place.

~~Development Program~~

~~Phase 1 Phase 2 Phase3~~

~~Fundamental Desigf no Design of Demonstration Unit Fabrication & Construction Utility Unit of Demonstration Unit~~

~~Fulfillmen f Componeno t t Demonstratio f Performanceno s & Design Technologies & Functions of Components~~

. Conclusion12 s Assumin gfull-scala e developmen f MHTGR-GTo t e authorth , s have completee dth conceptual desigrecuperatee th f no d cycle intercooled heliu turbins closeme ga th f edo cycln eo the basis of MHI's own technologies and various other information currently available. As a result, study results nearly identical to the results already obtained in the U.S.A. and other countries have been obtaine a potentia a mord r an de fo l compact desig s alsha no been confirmed. Furthermore, developmental problems regardin respective gth e componente th d san total system have also been clarified and the development program up to the stage of practical thif turbins o sga e bees us eha n summarized futuren I . intene w , realizo dt Modulae eth r High- Temperature Gas-cooled Reactor in cooperation with Japan Atomic Energy Research Institute and other organizations concerned as well as manufacturers in this field both domestic and abroad.

~~ACKNOWLEDGEMENTS~~

The authors wish to thank Prof. L.M. Lidsky and X.L. Yan at MTT and Mr. W.A. Simon at GA and other people who kindly assisted us by providing us with their valuable advic varioud ean s useful dat r thiafo s researc developmenh& t work sincerr Ou . e thanke sar also due to Dr. T. Yuhara who supported us in planning this research work and to Dr. I. Matsumoto, Mr. S. Morohoshi and Mr. S. Morii who took charge of various computational requirements. REFERENCES ] DOE[1 , 1986, "Concept Description Repor - Referenct e Modular High Temperature Gas- Cooled Reactor Plant", DOE-HTGR-86-118, U.S. Departmen Energyf o t . [2] AiResearch [3] Yan, X.L. and L.M. Lidsky, 1991b, "Design Study for an MHTGR Gas Turbine Power Plant", Proceedings of 53rd American Power Conference, Chicago, 111., April 29 - May 31. [4] "Evaluation of the Gas Turbine Modular Helium Reactor", DOE-HTGR-90380, December 1993. [5] C.F. McDonald, F.A. Silady, R.M. Wright, K.F. Kretzinger and R.C. Haubert,"GT-MHR Helium Gas Turbine Power Conversion System Design and Development", GA-A21617, GA Project 9819, March 1994. [6] Malcolm P.La Bar and Walter A. Simon, "Comparative Economics of the GT-MHR and Power Generation Alternatives".

108 [7] F.A. Siiady, AJ. Neylan and W.A. Simon, "Design Status of the Gas Turbine Modular Helium Reactor (GT-MHR)", GA-A21768, GA Project 9866, June 1994. [8] GCRA 1990, "Utility Projections of Operation and Maintenance Costs for Modular HTGR Plants", GCRA 90-005, Gas-Cooled Reactor Associates. ] McDonald[9 , C.F., 1992, "The e Closed-CyclFuturth f o e s Turbin Ga eRealisti A - e c Assessment", Proceedings of 27th Lntersociety Energy Conversion Engineering Conference Diegon Sa , , Calif., August 3-7.

~~KEXTP'• lefl.'-l INVESTIGATIO GT-SF NO T COMBINED CYCLE XA9642783 IN HTR-10 REACTOR~~

~~Z GAO, Z ZHANG, D WANG Institute of Nuclear Energy Technology, Tsinghua University, Beying, China~~

~~Abstract~~

In the Chinese HTR-10 project, two phases of the high temperature heat utilization are planned Durin secone gth d phase reactoe th , r core outlet temperatur increases ei p du o a 950°GT-St d Can T combined cycl s planei e econstructeb o t d d This paper present investigatioe sth n result GT-Sf so T combined cycl INEn patterno ei Tw T f so GT-ST cycles parallea e i , l GT-ST a seriecycl d san e GT-ST cycle referree ar , d Based on the state-of-the-art technology, the current activities of HTR-10 GT-ST combined cycle focus on a parallel combined cycle in which GT and ST cycle are independently paralle secondare th n i l y side

~~Keywords HTGR, Reactor Turbins Ga , e~~

~~1. Introduction~~

Ther currentls ei increasen ya d worldwid generatiow e ne interes e th f nucleano n i t r power plants which have extremely safe features, small size, technical and economical benefits Module High Temperature Gas Cooled Reactor(MHTGR) becomes new power supplienexe th tr centurfo r y s excellenbaseit n do t safety features, higs hga temperature and high thermal efficiency by combining the MHTGR and gas turbine technology

MHTG a passively-saf s Ri e nuclear reacto s expressei t I ra larg s a de negative temperature coefficient which naturally shuts the reactor down during heat up accident and is inherent safe in the reactivity transient The reactor low power density and huge heat absorbed capacity of the graphite structure assure that the decay heat will be dissipated passively by heat conduction and heat radiation even under the most severe accident condition MHTGR has also capability to produce very high helium outlet combinatioe Th s turbinC ga ° MHTGd temperaturf 0 - o an ne 95 GT o R( t p u e MHTGR) represents the ultimate in safety and economy. The high temperature heat source produce MHTGy db e use b o providdt n Rca e high thermal efficience Th y direct gas turbine (GT) cycle with MHTGR could enhance system thermal efficiency up to 50 % The high temperature helium coming from MHTGR drives the gas turbine cycle directl e applicatioTh y f compaco n t plate-fin recuperator boost e thermath s l efficiency

Considering the problems of the possible radioactivity deposition in the turbine blades in the direct GT-MHTGR, and the lower inlet helium temperature (250°C -300°C) in e indirecth s turbinga t e cycle MHTGR indirece th , s turbinga t stead an e m turbine combined cycles (GT-ST-MHTGR) were studied The GT-ST-MHTGR can be used to generate electricity more efficientl reactoe Th y r hea transferres i t e secondarth o dt y

111 loop meany Intermediatn sb a f so e Heat eXchanger (MXSteaa d m)an Generator <$G) thin I . s paper patterno tw , f GT-Sso T cycles, i.e .parallea l GT-ST a cycl d ean series GT-ST cycle, are referred. In the former pattern, IHX and SG are connected wit reactoe hth r primary loop orderlyturbins ga e steae th , cyclth md eturbinan e cycle arrangee ar parallen di l mannerturbins lattee ga th e en rth I . patterncycl& e X ear IH , arranged in the secondary loop, while the SG & the steam turbine cycle are arranged in the third loop. The system thermal efficiencies of a parallel GT-ST cycle and a series GT-ST cycle are 47% and 45.6 % respectively.

Chinn I , astudie progresn i e ar s o develot s pa standar d MHTGR plant designA . lOMWth Test Module(HTR-10) for investigation of MHTGR is under construction at site of Institute of Nuclear Energy Technology of Tsinghua University (INET). There are two phases of the high temperature heat utilization in the project. In the first phase, reactoe th r core outlet temperatur 700°s econventionai a d Can turbinT S l e cycls ei secondare useth n di y secone loopth n I .d phase reactoe th , r core outlet temperaturs ei increased up to 950°C and a GT-ST combined cycle is planed to be constructed.

Based on the state-of-the-art technology, the current activities of HTR-10 GT-ST combined cycle independentln focua n o s y parallel Combined Cycld whicn an ei T hG ST cycl independentle ear y paralle secondare th n i l y side. This selected configuration will make full use of the current MHTGR design with a relatively low core inlet temperature (i.e.,250~300°C), change smoothly from the ST cycle in the first phase to GT-Se th T seconcycle th n ei d phase conventionaA . l use s helica i system e dX th lIH .

~~2. GT-ST combined cycle for a modular high temperature gas cooled reactor~~

~~Two kinds of indirect GT-ST cycle are investigated based on Siemens 200 MW pebble-bed modular high temperature gas cooled reactor(HTR-Module). Fig. 1 shows the HTR-Module and its primary loop.~~

~~2.1. Series GT-ST combined cycle~~

Figure 2 gives the series GT-ST-MHTGR system. Two units of the 200 MW HTR- Module reactor usee sar powes da r sources secone th turbins n I .d ga e cycle, nitrogen is use s workina d g fluid. Tabl give1 e e maith s n thermodynamic parameterse Th . reactor heat is transferred to the secondary loops by means of an IHX and then transferred to the water in the third ST cycle. The gas turbine consists of high pressure turbine and lower pressure turbine. The high pressure turbine and compressor are mounte singla n do e shafttotae t electriTh ne .l c MWepowe2 18 s i r, includins gga d steaan turbin me turbineMW outpu8 e8 outpuf o t f 109MWeo t e busbaTh . r efficienc 45.6s yi orde n I %. optimizo rt e system effectsystee e th th , f mo s parameters thermae onth l efficienc investigatede ar y . Results show that increasing reactor outlet helium temperature will boost system efficiency s lesi t si thaf I . n 800°C, combined GT-ST cycle will lose its advantage.

~~2.2. Parallel GT-ST combined cycle~~

Figure 3 gives the parallel GT-ST-MHTGR system. The difference is that the GT cycle and ST are parallelly arranged in the second side. The high temperature helium is partly used for the GT cycle, and the rest for ST cycle. The thermodynamic data are shown in

~~112 Pebble-Bed Core RPV~~

~~Gas Duct Vessel Blower~~

~~SG Pressure Vessel~~

~~Fig. 1 Flow Diagram of Steam Supply System of the 200MW HTR~~

~~Tabl Parameter. e1 GT-Sf so T combined cycle~~

parameter series parallel Reactor thermal power (MWt) 2x200 2x200 Heat loss (MW) 5.6 5.6 Power output of gas turbine (MWe) 88 93 Power outpu steaf o t m turbine (MWe) 109 110 Power for plant use (MWe) 10.9 11.0 Power los generatof so r (MWe) 3.7 4.0 Net output power (MWe) 182.4 188.0 t systeNe m efficienc) y(% 45.6 47.0 Pressure ratio of gas turbine 3.5 5.1 Reactor power density (MW/m3) 3.0 3.0 Reactor coolant temperature (°C,inlet/outlet) 317/950 320/950 Helium pressure (MPa) 6.0 6.0

~~113 900/6 0 950/6 0 GTH~~

~~88MW~~

~~200M~~

~~588/1 83~~

~~-5 37MW~~

IHX-inlemiediate heat exchanger Gl H-high pressure turbine GTL-low pressure turbine GC-compressor ST-sleam turbine SG-steam generator Fi g2 Configuratio f no Scrie s GT/ST Combined Cycl 200Mr fo e W MHTR

~~950/6 0 925/6 0 555 Ikg/s N,~~

~~689/Z58~~

~~93MW GTL~~

~~88 83kg/» H,0 -5~~

Q-IX-intennedtale heat exchanger GUI-high pressure turbine GTL-low pressure turbine GC-compressor ST-steam turbine SG-steam generator PH-preheater Fi g3 Configuratio f no Paralle l GT/ST Combined Cycl 200Mr fo e W MHTR

114 1 core cavity 7 hot gas duct pressure vessel 2 pebbld ebe 8 internal tube of hot gas duct 3 cold helium hole 9 core support structure 4 hot helium hole e circulato 0 inleH 1 f o t r 5 hot helium plenum 1 outlee circulato1 H f o t r 6 annular space between reactor 12 annular space betwee pressure th n e pressure vesselan e cordth e vessel and the sleeve sleeve 3 stea1 m generator

~~Fig 4 The Cross-Section of the HTR-10 Primary Circuit~~

totae Tablt electriTh ne l e1 c poweMWe8 18 ,s ri includin turbins ga e geth outpu3 9 t MWe and steam turbine output 110 MWe The busbar efficiency is 47% The higher power efficiency and smaller IHX are the advantages of the parallel GT-ST cycle But on the other side, the water ingress accident will be caused probably by the rupture of the steam generator tubes

~~. GT-S3 T combined cycl HTR-1f eo 0~~

HTR-10 uses spherical fuel elements with ceramic coated particles, graphite th s a e core structure materia heliud coolane an th lfue e s ma lTh t element chargee sar d from the core top and removed from the core bottom via a discharge tube with multi-pass recycling Fig 4 shows the cross section and primary circuit of the HTR-10 The primary system consist reactoa f so r pressure vesse IHX-Sd an l G vesse arrangee ar l n di

115 reactor confinement cavity sid sidconnectey d eb ean coaxiaa y d b ducs lga t vessele Th . reactor core, reflector, carbon brick, control rods, thermal shielding as well as absorber ball syste e arrange mar e reacto th n i dr pressure vessel e intermediatTh . e heat exchanger and steam generator are included in the EHX-SG vessel. The steam generator consist smal7 3 f lo s coil pipes isolated with each othe located e an rth n i s outer annular region of IHX-SG vessel. The helical tubes of IHX are located in the central region heliue Th . m flows throug reactoe hth r cor downwarn ei d direction then through hot gas duct to IHX in upward direction and SG in downward way, then flows into blower cole Th d. heliu mpumpes i d from oute s annulaga r r duct intreactore oth . firse Inth t phas HTR-10f eo reactoe th , r operate outlea t sa t temperatur, °C 0 70 f eo cyclG S onluse s e i e power yth dfo r generation secone th n I .d phase reactoe th , r outlet temperature reaches 900°C. Several kind GT-Sf so T cycl studiede ear , including series GT-ST cycle, parallel GT-ST cycle and independent parallel combined cycle.

~~3.1, Ideal seria paralled an l l GT-ST combined cycl HTR-1f eo 0~~

Fig. Fig.d 5an 6 sho schematice wth f GT-Sso T combined cycle principlen I . ideae th , l seria paralled an l l GT-S HTR1f To 0 combined cycl t similaea r backgroun GT-Ss da T of MHTGR ,i.e., the same core inlet and outlet temperature and similar system , also cat hignge h thermacorrespondin% 43 d l an efficienc % g41 f .Tho y e main parameters showe ar tabln i . e2

~~930/3.0 14.9Kg/sN2 900/3.2 435/3.43 GT ST 2.84Kg/s 2.84 2.96 3.48 H2O MW MW HE~~

~~IHX HTR-10TM 10 MW~~

~~GC SG 41.5/ .008 -2.20 8.9MW 300/ 287/ MW 3.051 2.9 41.5/.008~~

~~-0.2MW 244/3.3 100/1.22 42/4.0~~

~~IlIX-intennediate heal exchanger GT-gas turbine GC-gas compressor SG-steam generator ST-steam turbine SC-steam condenser Fig.5 Configuratio f no Serie s GT/ST Combined Cycl r HTR-10Tfo e M~~

~~116 950/30 n.7Kg/s N2 900/3.0~~

~~GC GT 296 31-J 462 731 •3.06MW Kgfs MW HI 7.14 MW ^H. IHX -v*-^, *-*" 60/0.554~~

~~10 • MW 4S6/295 41 3/008 41J/.008 42/40 ST \!) 43J5/3.43 M sc 301 76M2 W SG MW PH 556 MW~~

~~300/ 287/ 230/ 516/ 30.r 2.9 3.9/065 2:'5 Kg/s H2O 0.594~~

~~-02MW~~

IIIX-mlermedialc heat exchange f-gaG s turbinr e GC-gas compressor SG-steam generator ST-stcam turbine SC-steum condenser PH-preheater Fig.6 Configuration of Parallel GT/ST Combined Cycle for HTR-10TM

~~Table 2. Parameters of combined GT-ST cycle of HTR-10 system~~

parameter series parallel Reactor thermal power (MW) 10 10 Thermal powe fflf ro X (MW) 10 7.14 Thermal powe(MWG S f ro ) 89 3.0 Output power of gas turbine (MW) 3.48 4.62 Output powe steaf ro m turbine (MW) 2.84 2.76 t outpuNe t power (MW) 4.1 4.3 t systeNe m efficienc) y(% 41 43 Reactor power density (MW/m3) 2.0 2.0 Reactor coolant temperature (°C,inlet/outlet) 300/950 300/950 Helium pressure (MPa) 3.0 3.0

~~3.2, Independent parallel GT-ST combined cycle of HTR-10~~

Base state-of-arn do t technolog consideratiod yan purposee th f no f botso h phasen si the HTR-10 project e independenth , t parallel GT-ST combined cycl f HTR-1eo s 0i more suitable. It will make reasonable utilization of HTR-10 heat resource with a

~~117 900/3.0 164/1 41~~

~~17V.E/1 1 S i , 483/3.1 Recuperator Reactor~~

~~10MW~~

~~300-3.05 Condenser~~

~~= 90%~~

~~blower 77 = 33%~~

~~Fig. HTR-1e 7Th 0 GT-ST Combined Cycle~~

~~Tabl . Parameter3 e f o sindependen t parallel combined GT-ST cycl f HTR-1o e 0 system~~

Contents 1 2 3 4 5 6 7 8 Reactor thermal power (MW) 10 10 10 10 10 10 10 10 Therma l(MW poweX IH ) f ro 5 5 5 5 5 5 5 5 Thermal power of SG (MW) 5 5 5 5 5 5 5 5 Working fluid of the GT cycle He He He He N2 N2 N2 N2 Inner efficiency of the gas turbine and 90 86 90 86 90 86 90 86 compressor (%) Net power output of gas turbine (MW) 2.01 1.73 1.92 1.61 2.08 1.78 1.94 1.63 Power output of steam turbine (MW) 1.36 1.36 1.36 1.36 1.36 1.36 1.36 1.36 Net power output (MW) 3.36 3.08 3.27 2.96 3.42 3.14 3.30 2.98 Net system efficiency (%) 33.6 30.8 32.7 29.6 34.2 31.4 33.0 29.8 Reactor coolant temperature (°C,inlet/ 300/ 3007 300/ 300/ 300/ 3007 3007 3007 outlet) 900 900 900 900 900 900 900 900 GT Nitrogen temperature (°C, inlet/outlet) 85Q/ 850/ 850/ 850/ 850/ 8507 8507 8507 550 550 500 500 550 550 500 500 Pressure ratio of GT 2.36 2.46 2.80 2.94 3.33 3.53 4.29 4.57 Pressure ratio of HC & LC 1.68 1.71 1.75 1.89 2.09 2.09 2.35 2.43 Helium pressure (MPa) 3.0 3.0 3.0 3.0 3.0 3.0 3.0 3.0

118 relatively low core inlet temperature (i.e.,250-300°C). The system power efficiency reaches acceptable value of 33%. Fig. 7 gives the schematic of independent parallel GI- ST combined cycle. The main parameters are listed in table 3. Table 3 shows that if the inner efficiency of GT & GC change from 90% to 86%, the efficiency loss is about 3%.

The current work associated wit secone hth d phase developmen f HTR-1o t 0 includes the desig f IHX-Sno G accident y vesselke e th , s analysi reactoe th f so r operatinp u n gi to 900°C~950°C outlet temperature. The license for the second phase must be applied futuree inth .

~~REFERENCES~~

(1) Lidsky, et al. Modular gas-cooled reactor gas turbine power plant designs , the second JAERI Symposium on HTGR Technologies , Tokyo ,Japan , October 21- 23, 1992 L.Yan) . Lidsky(2 M . L , , MGR-GTI indirecn A . t close turbins dga cyclR e poweeMG r plant for near-term development, MITNPI-TR-040, MIT, Aug. 1991 (3) Wang Dazhong, Zhang Zuoyi s turbine/steaga , e Studth f yo m turbine combined a modulacycl n o e r high temperatur s coolega e d reactor. Journa f Tsinghuo l a University , No.S333 l vo , 1993, pl-8 (4) Tong Yunxian, Wang Dazhong, Gao Zuying, Comprehensive study on JJTGR GT- ST combined cycles, Journal of Tsinghua University, Vol. 34, No. ES2 (English), 1994

~~119 HTG TURBINE-MODULS RGA E DEMONSTRATION TEST -A PROPOSAL XA9642784~~

~~E. TAKADA, K. OHASffl, H. HAYAKAWA, O. KAWATA, O. KOBAYASffl Fuji Electric Co., Ltd, Kawasaki, Japan~~

~~Abstract~~

The concept of HTGR (High Temperature Gas-Cooled Reactor) plant with Closed Cycle Gas turbine is taking an increasing interest because of its unique characteristics suc highls ha y excellent safet higd yan h thermal efficiency which reaches abou percents0 5 t necessite Th e demonstratio. th f yo nfule th tesl f o t scale power conversion modul usee b thin di o et s plan s beeha t n pointed out. This paper presents a new proposal of demonstration test facility concept.

~~1. Introduction~~

As for the module type high temperature gas-cooled reactor (HTGR), the range of application as the thermal energy source is very wide because the primary coolant temperatur reactoe th t ea r outle mucs i t h higher compared with other types of reactors. Therefore, utilization plans in various fields have been examined. Among these, Direct Cycle Gas Turbine Generation Plant recently attracts attentio nexe th ts nstaga e power generatio s excellenit n o t plan e tdu t thermal efficiency. Now, power generation plant named GT-MHR; which consists of 600MWth HTGR connected with Power Conversion Module (closed cycle regenerative gas turbine system contained withi nsingla e vessel undes )i r developmen. US e th y tb and Russian cooperation. Necessity of The Integrated Test Facility which verifies performance of Power Conversion Modul s beeeha n pointe t froe viedou mth w point that before th e Module is incorporated to the reactor system, its integrity and soundness of structura completelf l o desig d an n, y assembled full scale hardware shoule db demonstrated under non-radioactive full temperature and full speed conditions.

121 turbins ga e eTh 2syste. testee b mo t d To establish the concept of the Test facility, the gas turbine system to be targee th aime s a t plant dexaminea s twa followin e th n di g way. Two types of closed cycle systems illustrated in Fig. 1 were considered, and thermal efficienc s surveyewa y r varioufo d s turbine inlet temperatured an s pressure ratios. Other parameters which affect thermal efficiency were kept constane th o t t value tabulated in Tab.I. As a result, parameters shown in Tab.I were selected. This system yields thermal efficienc unde% 50 rf yturbino e inlet temperatur 850°Cf eo . In the following study, thermal output of the reactor will be assumed to be 200MWth. This value was selected considering that choosing somewhat smaller unit outpu expecd an t t mass-productio turbins ga n e effeceth f systeo t m woule db the more realistic and faster way to realize gas turbine HTGR.

3. Concept study of the Test Facility 1 Examinatio3. n pressure ofth e level Closed cycl turbins ega e syste featura s mha e tha controlliny tb pressure gth e level of the loop, the same thermodynamic condition as that at the design rated retainee poinb n ca t d even under partial load condition temperaturee th f I . t sa each cycle poinsame kepe th th f ear o eo t t value thoss sa t desigea n condition, pressure ratio and thermal efficiency are maintained .while shaft-end output varies in proportion to loop the pressure.

~~) (1 Powe c p/p° r * here, Power : Turbine shaft-end power output P : Fluid pressure Flui: P*d pressur reat ea l machine design point~~

From this relation, thermodynamic performance of the system can be verified by operating it under lower pressure level than the design pressure. When such a way was chosen, the capacity of the heater used instead of reactor and the heat made b sin n ek ca muc h smalle proportion ri P/P*o nt .

~~Q(H)orQ(S) oc p/p* (2) here, Q(H): Thermal power of heater (nuclear reactor) Q(S Capabilit: ) heaf yo t sink~~

~~122 05~~

~~.0. Reactor ra *->~~

~~Power Conversion Module Entropy "S" [kJ/kg • °C] cw Intercoolen No ) (1 d Regenerative Cycle~~

~~o> _^ 3~~

~~Reactor re *-> C P i r~~

~~Power Conversion Module Entropy "S" [kJ/kg cw (2) 1-Stage Intercooled Regenerative Cycle~~

~~Fig.1 Two Types of Closed Cycle Systems Studie Plane th tdn i Selectio n~~

123 TABLE L MAJOR PARAMETER OF TARGET PLANT Reactor (Heater) Power MWt 200 Compressor inlet pressure MPa 2.86 Compressor outlet pressure MPa 8.0 Cycle Pressure ratio — 2.80 Compressor inlet temp. °C 30 Turbine inlet temp. °C 850 Turbine adiabatic efficiency — 0.91 Compressor adiabatic efficiency — 0.91 Recuperator efficiency — 0.95 Numbe f Intercoolinro g — 1 Turbine shaft-end power output MW 102 Thermal Efficienc t shaft-enya d % 50.8

Although above expressions are theoretically correct, there are actually some essential or inevitable losses such as mechanical loss, auxiliary loss, and etc. ,which can not be neglected in extremely small partial load. Therefore it is necessar examinyo t whao et t poin reducede b t P/Pn ca * . The change of the thermodynamic characteristics due to relative pressure level, (X=P/P*) can be shown in Fig.2.

~~Here, X=P/P* (relative pressure level) Y=Power/Power* (relative turbine shaft-end output) rj I T) *=y/x (relative thermal efficiency) Power ; Turbine shaft-end output (* : design point value)~~

~~Fig.2 suggests if inevitable losses of the system can be assumed 3% or less, X=0. morr 2o e seems sufficien predico t t t full power efficienc systeme th f yo .~~

3.2 Flow Diagra Energd man y Balanc Tese th f te o Facilit y When the relative pressure level X=0.2 is adopted, required heat input to the Test Facility can be reduced to 40MW ;20% of the target plant thermal output. In addition, when considerin turbins gga tha e eth t system thermaa itsel s ha f l efficiency of 50%, a half of the required heat input can be recovered by the turbo generator turbineconnectes ga e th . o d t sho 4 Fig. floe d wth 3w an diagrad man

~~124 "-L Define SJ yi= TJ I T] * : relative thermal efficiency O y=power/power* : relative power~~

~~equation y2=ax-b where a=l+b b=inevitable loss ratio~~

~~b= 4 0. 2 0. 0 ' 0.6 0.8 1 relative pressure level X(=P/P*)~~

~~Fig.2 Effec "inevitablf to e loss Rean "i l Plant Partial load Characteristic Closef so d G.T.~~

...... Turbine output powe. .. r . recirculatio.^ n [20MWe] C.w. i\f I.e. A V S.OMPa.-SSOt:, 21.8ke Is A $ 1 ^ 2.86MPa, 30*C i 3 • --^r^^ iI \f 1 ••^"""^ j T fo 2 I 4 520X: lost: Electric | \f \ ' 1st2; / Heater / ^ | LAVV\r A ' 6 ^) ———— 5 ! -AAA /V •••••^^ ^ P.C. R.X. • Pow r Conversioe n Module From outside electric power grid [20MWe] c.w.

~~Fig.3 Flow diagram of Integrated Test Facility~~

~~125 Power from outside Power Pre- Heat Release Conversion Cooler (20MWe) Electric (20MWt) Module Heater Gener (40MWt) -ator~~

~~Electric power output recycle (20MWe)~~

Major Specification of Test Facility Heater Power Output 40MWt Compressor inlet pressure 0.57MPa Compressor outlet pressure 1.6MPa Cycle Pressure ratio 2.80 Compressor inlet temperature 30.0°C Turbine inlet temperature 850°C Turbine adiabatic efficiency 0.91 Compressor adiabatic efficiency 0.91 Recuperator efficiency 0.95 Number of Interceding 1 Turbine shaft-end power output 20.3MW Thermal Efficiency at shaft-end 50.8%

~~Fia.4 Energy balance of integrated Test Facility~~

126 energy balance of this method. 20MWe electric power from outside and 20MWe fro turbe mth o generato suppliee ar rtese th t o dloot p electric heater thesef O . , 20MW is recovered by the turbo generator of the power conversion module and the rest are released to the environment. Electric heater was preferred from the view point that it is the easiest way to recover turbine output, and when compared with fossil fired high temperature heate provee seemt th r i e b n o stechnologt experiencer ou n y i als s i t oI expecte. d that by using electric heating, various test conditions including that for load variation test can be controlled more easily than in case of fossil fired heating.

3.3 Test Items By using this facility addition i , performance th o nt e verification under partial load, but full temperature and full speed condition, the following tests can be performed in a clean helium environment prior to connect the nuclear reactor. (D Response systee th f so m under various transient conditions Response of the system under startup, shutdown operation, ramp or step chang loaf eo d ,los electricaf so l grid load ,los coolinf so g water supply etc. will be tested. (2) Flow distribution measurement Flow distributio varioun ni s flow paths suc thoss ha e among recuperator banks, at inlet and outlet of the turbo machines etc. will be measured. (3) Interface problem resolution If the interfaces between each components are proper, if they are not e damagedunexpecteth f i r o , d leak path d flow e formean sar s y b d differential thermal expansio d othean n r operating loads, etc. wile b l verified. (4) Maintenability verification Methods and remote handling equipments (if necessary) for removal or installatio turbo-machinef no testede b n sca . (5) Operator training

. 4 Conclusion It is said that no fatal problems are foreseen in developing each components of the Power Conversion Module such as turbo-machinary, magnetic bearings, high performance recuperator, and etc. However, for the complex system like this, it is absolutely necessar understano yt e characteristicdth e entirth f eo s systey mb assembling and operating them prior to deployment as a nuclear plant system. The Integrated Test Facility proposed in this paper will provide realistic and economical way to meet such a requirement.

~~127 LICENSING, FUEL AND FISSION PRODUCT BEHAVIOUR~~

~~Session3 A STUDY OF SILVER BEHAVIOR IN GAS-TURBINE XA9642785 HIGH TEMPERATURE GAS-COOLED REACTOR~~

~~K SAWA S SfflOZAWA, KUNITOMIK , , T TANAKA Department of HTTR Project, Japan Atomic Energy Research Institute, Ibaraki, Japan~~

~~Abstract~~

In High Temperature Gas-cooled Reactors (HTGRs), some amounts of fission products (FPs) releasedare from fuel elements transportedare primaryand the in circuit -with primary coolant during normal operation. Condensable FPs plateout on the inner surface of pipings and components in the primary circuit and the gamma-ray emitted from the plateout FPs becomes main source during maintenance works. In the design of a Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR), behavior of FPs, especially of silver, consideredis important from the view point of maintenance works of the gas-turbine. However, the behavior of silver is not well known comparing with that of noble gases, iodine and cesium. Then a study of silver behavior GT-HTGRthe in carriedwas basedout experienceson of High Temperature engineering Test Reactor (HTTR) design. purposesThe of this study determinewereto (1) how important the silver in the GT-HTGR, (2) to find out countermeasures to prevent silver release from fuel determine to elements (3) itemsthe and future of research developmentand which will be needed. thisIn study, evaluations of (I) inventory, (2) fractional release from fuel elements, (3) plateout distribution primarythe in radiation circuit(4) and gas-turbine dosethe at were carried out. Based thison predicted study,was it thatgamma-raythe from plateout silver contributes about halfa of total radiation dose during maintenance workturbinegas ofIn the future, expectedis it that more detail data of silver release from fuel, plateout behavior, etc. will be accumulated through HTTRthe operation.

1. INTRODUCTION In high temperature gas-cooled reactors (HTGRs), some amounts of fission products (FPs releasee ar ) d mainly from fuel with coating transportee ar d primare san th o dt y cooling system wit primare hth y coolant during normal operation than I . t case, condensabl plateous e eFP th n to inner surfac pipingf eo componentd san primare th n si y cooling system othee th n r handO . , (1)>(2) sinc HTGRe eth heliue sus m gas, t activatewhicno s hi d itself s primara , y coolant, almoso n t amount of corrosion products is generated. Then, the gamma-ray emitted from the FPs becomes main source in shielding design of the primary cooling system of HTGRs. In order to calculate the

131 plateout distribution in the primary cooling system of HTGRs, a computer code, PLAIN, has been developed applicabilite th d (2) desig e an cod,e th th bees o ef t n yha o n examined(3)'t4). Becaus f higeo h diffusion coefficient coatinn i s g layer graphited san emittind (5an ) g relatively high energy gamma-ray, n0mAg would be important in the shielding design and planning of maintenance work of gas-turbine HTGRs (GT-HTGR)(6)'(7). Sufficient plateout distribution has t beeno n observe plateoun di t experiment t JAERsa I because activation a A s gi n producf o t 110m 109Ag whic vers hha y small yiel uraniuy db m fission (two orders smaller than tha 131f o t137r Io Cs). A preliminary study of silver behavior in the GT-HTGR was carried out based on experiences of High Temperature engineering Test Reactor (HTTR) design. The purposes of this study were (1) to determine how important the silver in the GT-HTGR, (2) to find out countermeasures to prevent silver release from fuel elements and (3) to determine the items of future research and development which will be needed. This paper describes evaluation results of (1) inventory, (2) fractional release from fuel elements, (3) plateout distribution in the primary circuit and (4) radiation dose at the gas-turbine.

. INVENTOR2 Y CALCULATION e characteristiTh c feature f importano s categorP F t HTGn i y R systeme b n ca s summarized below. (1) Noble gases These can be retained by intact coating layers, then these are released from failed particles. Noble gases exist in the gas phase and can be removed by primary coolant purification system. Though some of them have condensable daughter FPs, this effect is not large at failure fraction of HTGR fuel. (2) Halogens These behavio approximatels ri y nobl e samth s eea gases with respec releaso t e from fuel. However, becaus modesa f eo t chemical affinit metalsr yfo , halogens ten chemisoro dt n bo solid surfaces. In this study, the behavior of 131I was investigated. ) (3 Alkali metals These have a higher chemical affinity for graphite than the halogens. The sorption is sufficiently strong such that graphite is an effective barrier to release from the core under normal operating conditions mose Th .t significant featur cesiuf eo m chemistry relativ plateouo et s it s i t

high affinity for oxides. These form on diffusion of cesium into the oxide protective layers on steel platee th x dtend fi materiao an dt placen i l year.0 Sinc3 half s Cso seha f 137 life plateous it , t amount is accumulate bees ha importann na d dan t nuclide fro viee mth w poin shieldinf o t g desige th f no HTGRs.

132 (4) Rare Metals Because of high diffusion coefficients in coating layers and graphite(3) and emitting relatively high energy gamma-ray, ll0mAg would be important in the shielding design of HTGRs. Though I10m activationn a A s gi product from I09Ag whic vers hha y small yiel uraniuy db m fission (two orders smaller than tha ljl f 137r largeto s lo Cs) ha t i ,r yiel plutoniuy db m fission meant I . s that low-enrichee inth d uranium cyclhigo t p heu burnup inventore th , f yo Ag becomes large. . 110in Tabl showe1 s fission yield half-lived san 131f so I , iy? 110md Csan Ag (including fission yields and absorption cross section of I09Ag). Inventories of I31I and 137Cs are calculated by the following equation.

~~Tabl e1 Compariso fissiof no n yiel half-lifd dan f eo I , "d Csan Ag . 13I !37 0in Nuclide Fission yield Half-life Cross section~~

~~13II 3.23% day1 8. s 137Cs 6.35% 30 years~~

~~l 9 235 ° Ag 0.0 ( 3% U) 00 93.5 barns 1.66%(239Pu)~~

~~I!0m "Aeo 0 252 days~~

~~— (2.1)~~

~~Where, N :number of nuclide (atoms), P :reactor power (W), Y : fission yield (-), X :decay constant (s-1). Inventory of mAg is calculated by the following equations. dN, 109 (2.2) dt~~

~~133 dN, 110m (2.3) dt~~

•10, 9 Where, N 109 :number of Ag (atoms), 110m N 110m :numbef ro Ag (atoms), a :fraction of plutonium fission, • 109 YPu plutoniuy b g A :yielm f d o fission , :yield of• 109 A, g by uranium fission, :capture cross section of 109Ag, :thermal neutron flux (m'Y1).

Figure 1 shows calculated result of 110mA, g inventory in GT-HTGR core. In the calculation, plutonium fission fractio thermad nan l flutreatee xar parameters da resule th d t san shows that inventory of Ag depends on plutonium fission fraction and thermal neutron flux. 110m Tabl indicatee2 s predicted core design parameter GT-HTGf so R together with thos HTTRf eo . With higher burnup, higher power density and longer irradiation time, plutonium fission fraction, thermal flux and irradiation time in GT-HTGR core could be 60 %, 4xl0 cm'V, 6 years, 14 respectively. Base Tabld Fign an d o 1 e. 2,110mAg inventor GT-HTGn yi R core, which thermal , wil aboue b l MW time0 powe7 t0 45 s s largei r r than tha HTTRf o t .

~~1E+7~~

~~> 1E+? 6~~

~~-.%r.-.-Q:t"&-"--Q"-&-- 1E+5 Thermal neutron flux O 4xlO"cm:s' 2x/0"c/77-A y Q 1*10" em's' 1E+4 0123456789 10 Irradiation time (EFPY)~~

~~Fig 1 . Calculated inventor 110mf yo Ag.~~

~~134 Table 2 Core design parameters for Ag inventory calculation. Parameter HTTR GT-HTGR Pu-fission fraction 30 60 Thermal flux (xl014cmV) 0.6 4 Irradiation time (EFPY) 3 6~~

3. RELEASE FROM FUEL ELEMENT The coating layers of fuel particles provide barriers to the release of iodine. Since short-live disintegrats dFP e during diffusion proces coatinn si gconsideree layersb n ca t ,i d that iodine is mainly released from fuel particles with defects in their coating layers (i.e. failed particle). A calculation mode bees lha n develope evaluato dt ratie ereleasth f oo birtho et , (R/B) f iodino , e from the fuel compact containing failed particles(5)'(8). The release process of the cesium and silver from the fuel rod can be described by two consecutive steps firse :th t release occurs from fuel particle thed soccurt an ni s fro fuee mth l rod. The former release is controlled by two physical mechanisms - diffusion and recoil - and the latter release, diffusion in the interior of the fuel compact and sorption/evaporation on the surface. In analysie th f fractionaso l releas f metallieo , threcFP e type f particleso considerede sar : intact, degraded and failed particles. Since the SiC layer acts as the primary barrier to the release of metallic FPs particle ,th e whic onls hha yfaile a layeC dSi r shoul alse db o modele addition di o nt the intac failed tan d coated particles. The degradee nth d particle correspond particle th o st e with failed SiC layer. Moreover, they can be released from the intact coated fuel particle which is irradiate t higheda r than about 130 durin0C ° g long period knows i t I . n that fractional releasf eo I10mAg is higher than that of 137Cs. Fractional release I31f so 1, 137 1!0md C san Ashowe gar functio a Fign ns i a . 2 failurf no e fraction. In the figure, X-axis shows through coating fraction and in the calculation, failure fraction of SiC layer is assumed to be 10 times of that of through coating failure fraction. Since fractional release of iodine is proportional to failure fraction, it can be reduced with minimizing the through coating failure fraction. The fractional release of cesium can be reduced as low as 10"*~ 10'5 by minimizing failure fraction, however, diffusive release from intact particle should be reduced to attain lower fractional release than 10"5. For silver, since diffusive release fraction from intact particle is higher than that of cesium, the fractional release does not depend on the failure fraction. This is a reason why silver could be taken into account in the GT-HTGR design. The countermeasures to decrease diffusive release from intact particle are, for example, ) (1 reduce fuel temperature (the maximum temperatur 110e< 0 °C), and/or ) (2 adoptio f diffusivno e resistant materia coatins a l g layer (ZrC layer).

~~135 1E+0~~

~~IE-6 IE-6 IE-5 IE-4 IE-3 IE-2 Failure fraction (Exposed failure)kernelSiC or~~

~~Fig. 2 Fractional release of fission products.~~

The fuel temperature reduction depends on core design. In order to attain higher thermal efficiency, reactor outlet gas temperature cannot be reduced and it is difficult to reduce the fuel temperature. The diffusion coefficient of 137Cs in ZrC layer is about two orders lower than that in SiC layer. For silver, smaller diffusion coefficient in ZrC layer than in SiC was obtained(9), however, sufficient data hav t beeeno n accumulated.

. PLATEOU4 T DISTRIBUTION computeA r code, PLAIN bees ha ,n develope JAERn di calculato t I plateouP eF t distribution in the primary cooling system of HTGRs . PLAIN is based on Iniotakis model but (2) (10) ha modifiesa penetratioP d F modee th n i l n process int base oth e metaP. verificatioThe n works were carriecomparinby dout calculategthe plateoudFP t distribution with experimental data obtaine Oara (OGL-1)n d i 1 Loos . iGa pNo (1), whic installes hi d Japae inth n Materials Testing Reactor (JMTR JAERn )i simulated Ian primare sth y cooling system conditio HTGRf nOGL-1o e th n I plateoue . th , t distributio bees nha n measured fro outside mth e of the primary pipes after every operational cycle using Ge-detector. The range of helium gas temperatur fros ei mt tesa t100 C fue° 0 l exi rooo t m temperature. Flow conditio turbulents ni . Flow diagra primarmf o y cooling syste OGL-mf o shows i 1 Fign ni . 3(I). Examples of verification results for 131I and 137Cs are shown in Fig. 4. In the figures, black circle lined san s show measure calculated dan d plateout densities, respectively maie Th n.

~~136 heat exchanger~~

~~in-pile tube~~

~~fuel exit~~

~~Fig.3 Flow diagra OGL-1f mo .~~

~~hot duct duct in-pile tube h<:at exchanger cooler [ 1 1~~

~~20 40 60 Distance from fuel exit (m)~~

~~Fig 4 . Measure calculated dan d plateout distributio OGL-1n ni .~~

137 Table 3 Main plateout condition of OGL-1. Parameter OGL-1 HTTR Materials Stainless steel Stainless steel Hastelloy-X (PWC heat transfer rube) Hastelloy-XR (Inner pipe, IHX heat transfer tube) Cr-Mo steel (Annulus of heat exchanger, RPV) Coolant temperature 950°ORT' 950-395 °C Wall temperature 950°C~RT' 950-160 °C Heliu pressurs mga e 3MPa 4MPa Helium gas velocity 10-6s m/ 0 20-4s 0m/ * Room Temperature

plateout conditio OGL-n ni 1shows i Tabln i wite3 h tha HTTRf to . Fro resultse mth , though about an order of difference is locally observed between measured and calculated values, the calculated plateout profiles show good consistency with the measured ones as a whole in spite of the complicated temperature distribution and flow diagram as shown in Fig. 3. Then, it is concluded that the analytical model and the physical constants are applicable to predict the plateout distributions in the primary cooling system of HTGRs. Sinc HTTe eth R take t higsou h temperature heliu heas e fro corme th ga mt th o e t exchangers accumulatn ca t i , e good operational dat predico at plateoue tth t distribution othef so r HTGR sGT-HTGRe sucth cooles s i h a t i thermae y d b Th an . , l outpuHTTe MW th 0 3 f o Rts i reactoe th t a r C inle° 5 t whic39 heliuf o hs mflowga s downward throug core e reactoe hth Th . r outlet coolant temperature is 850 °C at the rated operation and 950 °C at high temperature test operation numbee Th . maif ro n cooling looheae oneth removes ps i i t d intermediat,n an a y db e helium/helium heat exchanger (MX pressurizea d )an d water cooler (PWC) loade paralleln di e Th . pressure of primary coolant is 4 MPa. Temperatures and flow rates of the components in the primary cooling system are shown in Table 4. In the table, values written in the parenthesis indicat operationC ° e 0 thos95 t ea . Plateout distributions in the parallel loaded operation at rated power operation, i.e. 850 °f outleCo t coolant temperature showpredictes e i ar ,t I Fign ni . 5 . d tha 137d t Can bots l hm plateout mainl heae th tn ytransfeo wherC r tubePW e f temperaturso primare relativels ei th n i yw ylo cooling system of the HTTR. On the other hand, there are very small amounts of plateout FPs on innee th r pip co-axiaf eo l double pipe where temperatur vers ei y hig therefored han , desorption rate is large by its high lattice excitation frequency. From the calculation results, it is predicted that

~~138 10'2 - Result of 131I plateout calculation -~~

~~\ \ PWC - 104 11 - .£ "s- j~~

~~| 1010 A~~

~~cs / L. ___ j— -_~~

~~3 / IHX o 9 x £ 10 — x - « - ——— —— •' E '~~

~~B 10 Inner Heot Tronsfer Pwc Outer RPV Pipe Tube Filter G/c ^ Pipe tah»~~

~~10" Result of 137Cs plateout calculation -~~

~~. "1 PWC I~~

~~w at /IHX -~~

~~10' Inner Heat Tronsfer Pwc Outer RPV ^ . , . IHr _ o X C Fi"e»...,i«G/ r . Pipe Tube t t « AnnuluP P j~~

~~Fig. 5 Predicted plateout distribution in HTTR.~~

139 Tabl e4 Main calculation paramete primarf ro y cooling syste HTTf mo R (Parallel loaded operation). Region Coolant temperature (°C) Wall temperature (°C) Flow rate (t/h) Inner pipe 850 (950) 850 (950) 44.6 (36.5) PWC heat 850-385 160-165 29.7(24.3) transfer tube (950-385) PWC CG/ 385-395 385-395 9.9* (8.1*) IHX heat 850-385 805-320 14.9(12.2) transfer tube (950-390) IHC XG/ 385-390 385-395 14.9(12.2) annuluV RP s 395 395 45.3 (36.5) () shows values in 950 C operation. * sum of 3 lines

cesium and iodine will plateout mainly on the wall of component whose temperature is relatively low (about 200-400 °C), in the concrete, the pre-cooler of GT-HTGR. The plateout distribution of silver was not observed in the OGL-1 experiment because the maximum fuel burnup in the experiment was about 5% FIMA and there was no enough silver inventor to measure its plateout distribution. Figure 6 shows plateout distribution of silver which was measure VAMPYR-Ie th n di I experiment measuree Th . d data show that silver tendo st 00 plateou relativele th n o t y high temperature wal 500-60f o l . Thoug0°C verificatioo hn bees nha n carried out, the calculated result of silver plateout distribution in HTTR by the PLAIN code is show Fign n i . Thi7 . s result shows that silver plateou hign to h temperatur . Thes°C e0 e wal85 f o l results suggest that silver will plateout on the gas-turbine in the GT-HTGR system.

5. RADIATION DOSE Radiation dose in the gas-turbine was evaluated based on these studies. The calculation codeD evaluatioe QA (12) carries y Th .b wa t dou n condition followss a e sar . (1) Inventory r """"A thermae Fo . Th g inventorlMW powe0 45 s yri calculation, plutonium fissio110mn, fraction and thermal flux are assumed to be 60% and 4* 10l" cm'Y1, respectively. (2) Fractional release from fuel The through coating and SiC coating failure fractions are assumed to be 4><10~5 and 1.6x10"*, respectively, which are one order lower than expected values in the HTTR. The maximum fuel temperature is assumed to be 1300 °C and SiC layer thickness is 35 um.

140 10 temperatur— e N <> experimental . . .. . — calculated with SPATRA. «*•" concentrato of E AQ-110m on Inconel 617 109 iHjjj^OkJ-moT1 tr activitx y mobilizey db o leachin waten gi r E ad/desorption equiDvium<[> mass transfer control o -1000

~~"5 107 -900 Z a i 106- a a. o 800 5 1 s c S ° o u E a u ao w w 123 Specimen length Im]——*-~~

~~Fig. 6 Plateout distribution in VAMPYR-II(11)~~

~~Heat| Exchanger\ | • IE-2h V ! "3 O l ~ IE-3 ——pwc, - - ///AT~~

~~IE-4 0 3 5 2 0 2 5 1 10 35 40 Distance from core exit (m)~~

~~Fig. 7 Calculated plateout distribution of Ag in HTTR.~~

~~141 Table 5 Result of dose rate. Nuclide Dose rate (fiSv/h) 137Cs 12 131I 5~~

~~110m AA g 125 Others 100 Total 242~~

(3) Plateout distribution The plateout fraction in the gas-turbine is the same as that in the inner pipe of the HTTR where temperature is about 850 °C, namely 0.03 % for I31I and 0.01 % for 137Cs (from Fig. 5). For UOmAg, it is assumed that all silver plateout on the gas-turbine components. (4) Radiation dose The radiation dose from gas-turbine depends on its detail design configuration, arrangement, shieldings thin ,I etcs . calculation assumes i t i , plateous d FP tha ) n (a t o t the inner surface of infinite tube, (b) an evaluation point is 1 m from the tube and (c) shieldin gf iron o materia m . c 3 s i l dose Th e rates from Cs, I, showe ar othed A resul e s Tablgan n ni rTh FP t. eshow5 s that 137 I3I 110m the total dose nea gas-turbine rth abous ei uSv/0 aboud t25 hhalan e thaf tth f o contributio s i t n from UOm'Ag.

6. CONCLUSIONS The calculated result of I10mAg inventory in GT-HTGR core showed that with higher burnup, higher power densit longed yan r irradiation time, plutonium fission fraction, thermal flux and irradiation time in GT-HTGR core could be about 70 times larger than that of HTTR. The fractional release of cesium from fuel can be reduced as low as 10^~10" by 5 minimizing failure fraction, however, diffusive release from intact particle shoul reducee db o dt attain lower fractional release than 10'5. For silver, since diffusive release fraction from intact particle is higher than that of cesium, the fractional release does not depend on the failure fraction. The countermeasures to decrease diffusive release from intact particle are reduction of fuel temperature (the maximum temperature < 1100 °C) and/or adoption of diffusive resistant material coatins a g layer (ZrC layer). It is predicted that both 131I and 137Cs plateout mainly on the heat transfer tubes where temperature is relatively low in the primary cooling system of HTGR. The measured data show

142 that silver tends to plateout on the relatively high temperature wall of 500-600 °C. The calculated result of silver plateout distribution in HTTR by the PLAIN code shows that silver plateout on high temperatur . Thes°C 0 ee wal85 result f o l s suggest that silver will plateou gas-turbine th n to n ei

~~the GT-HTGR system~~

~~m 110m The dose rates froACsmgl, 137 , and other FPs shows that the total dose near the gas-turbin abous ei uSv/0 about25 d hhalan e th tcontributios fi n from 110niAg.~~

~~ACKNOWLEDGEMENTS~~

The authors wish to express their gratitude to Dr. O.Baba of JAERI for their useful comments and support of this study. The authors are also grateful to many of the staff in the JMTR supporo wh , t measuremen plateouf to t distributio OGL-1e th n ni .

~~REFERENCES~~

1. TSUYUZAKI.N., MATSUMOTO,M.: JAERI-M 88-225, (in Japanese), (1988). 2. BABA,O., TSUYUZAKLK, SAWA,K.: JAERI-M 88-266, (in Japanese), (1989). 3. SAWA,K., MURATAJ., SAIKUSA,A, SHINDO,R., et.al.: J. Nucl. Sci. Technol., 31,654-661, (1994). 4. SAWA,K., BABA,O.: JAERI-M 91-084, (in Japanese), (1991). 5. SAWA,K., SfflOZAWA,S., FUKUDA,K., ICHJHASHI,Y.: J. Nucl. Sci. Technol., 29, 842-850, (1992). 6. WICHNER,R.P.:NUREG/CR-5647, (1991). 7. SAWA,K., TANAKAJ.: JAERI-Research 95-071 (in Japanese) (1995). 8. SAWA,K., FUJD,S., SHIOZAWA,S., HIRANO,M.: JAERI-M 88-258, (in Japanese), (1988). 9. CHERNIKOV,A.S., et.al.: IAEAIWGGCR/13, 17-0181, (1986). 10. INIOTAKIS.N., MAUNOWSKIJ., M NUNCHOW,K.: Nucl. Eng. Des., 34, 169, (1975). 11. MOOREMANN,R.,: JUL-2668, (1992). 12. CADSF,V.R.: RSK CCC-307, (1977).

~~143 XA9642786 COMEDI EXPERIMENT1 EBD : FISSION PRODUCT BEHAVIOUR DURING DEPRESSURIZATION TRANSIENTS~~

~~R. GILLET, D. BRENET DTP/SECC, CEA/GRENOBLE, Grenoble, France~~

~~D.L. HANSON General Atomics, San Diego, Calofomia, USA~~

~~O.F. KIMBALL ORNL Ridgek Oa , , Tennesse, USA~~

~~Abstract~~

experimentan A COMEDIA l prograCE e th E mn i loo bees pha n carrie obtaio t t dnou integral test dat validato at e methodeth transpord an s t models use predico dt t fission product release fro core plate-oud mth e an primare th n i t y coolant circuiModulae th f o t r High Temperatur Cooles eGa d Reactor (MHTGR) during normal operatio liftoffd d nan an , during rapid depressurization transients.

looe pTh consist in-piln a f so e section wit fuee hth l element, deposition section (heat exchanger), filters for collecting condensible Fission Products (FP) during depressurization tests and an out-of-pile section devoted to chemical composition control of the gas and on- line analysis of gaseous FP.

After steady state irradiation, the loop was subjected to a series of in-situ blowdowns at shear ratios (ratio of the wall shear stress during blowdown to that during steady state operation) ranging 5.6 o frot . 7 m0.

~~The results regarding the FP profiles on the plate-out section, before and after blowdowns are given. It appears that:~~

• the plate-out profiles depend on the FP chemistry depressurizatioe th • n phases significano t hav d ele t desorptioe th n I o 131 f t no bu , contrary, they have almost no effect for the other FP such as Ag 110m, Cs 134, Cs 137 and Te 132.

~~1. Introduction~~

The COMEDIE BD1 experiment to support the MHTGR (Modular High Temperature Gas cooled Reactor) design has been performed in the COMEDIE loop located in the SILOE GrenoblreactoN CE t a r e durin perioe gth f Septdo throug2 e 9 . th r hfo Nov2 9 . irradiation phase and December 92 - June 93 for the post irradiation examination.

This experimental program has been carried out in the framework of a US/CEA contrac managed RIDGK an t E OA sponsore Ey dDO b NATIONAS U y db L LABORATORY. Its goal was to give measured data to GENERAL ATOMICS to validate computer codes developed for the design of MHTGR reactors.

145 The primary objective of the BD-1 test is to obtain representative data on the release, transport, plateout liftofd f condensiblan ,o f e fission product in-piln a n si e test loop under nominally "clean" conditions. These data will the usee nb validatdo t e thadesige th t n methods use o predict d t fission product transpor e MHTGth n i t R hav e requireth e d predictive accuracies. Typically, these transport codes contain multiple component models which are derived from differential single effects tests performed in the laboratory or in-pile experiments. The purpose of these in-pile loop tests is not to provide fundamental data from which transport derivemodele b y t rathedsbu ma provido rt e integral test dat asseso at e sth validity of these integral computer codes.

~~The primary test objectives can be divided into three parts :~~

~~1. To obtain data on fission product release from a fuel element with a known fuel failure fraction particulan i , confiro rt effecte mth f higso h pressur higd ehan flo releasen wo .~~

~~2. To obtain data on plateout in an in-pile loop under nominally "clean" conditions expected during normal operation.~~

~~obtaio T 3.n dat fission ao n product liftoff ove rangra f sheaeo r ratios~~

~~2. Description of the COMEDIE loop :~~

~~COMEDIe Th E loop compose(figs i ) 1 .out-pil n in-piln a a d f do e ean sections.~~

~~2.1. In-pile section:~~

~~This section consists of:~~

fuee th l elemen • sourc e whicP tth F (fig s f i he o ) .wer2 e entraine coolane th s y db ga t after release fro particlee mth diffusiod san n throug graphitee hth . Some "designed to fail" (DTP) particles (LEU UCO kernels with a 23 urn pyrocarbon seal coat) which were expected to fail along the first cycle, seeded in fuel compacts with TRISO-coated particles.

• The H-451 graphite reflector block simulates the lower (core exit) reflector of a MHTGR determinn ca t i d depositioe ean th condensiblf no graphitn o P eF e structures. plate-oue Th t• section r heao (fig , t 3) .exchanger , consist f threo s e parallel tube three th bundles f eo bundlee isolates On . swa dblowdowe prioth o t r s i t ni tesd an t considere referencea s da .

~~• Four cartridge filters (fig. 4-5) which collected FP during the depressurization tests. A different cartridge filter is used for each depressurization test.~~

~~• 2 probes located on the upstream side and on the downstream side of the heat exchanger make possible sampling of coolant gas.~~

~~electrican A • l helium heate fue e temperatur s inle th s adjusl o ga t rn ga i t e e th t th t ea element.~~

~~146 ons /WING II? CO H20(VATOO)~~

~~FIG 1 - COMEDIE LOOP CIRCUIT DIAGRAM BUNDLE CO DEPOSIT SECTION~~

~~BUNDLES~~

~~DOWNSTREAM STOP VALVE~~

~~FLOW PATH DIAGRAI~~

~~6 HEAT EXCHANGER TUBES 48/10 • I CONNECTION TUBE w u~~

~~f E~~

~~•1390 idili~~

UPSTREAM STOP VALVE •1310 FUEL HOLES flnmn» THERMOCOUPLES 1 COMPACTS Jh;.5 Length 5Pit*

~~COMEDI- 2 G FUE REFLECTO1 FI D EBD L AN R ELEMENTS FIG 3 - COMEDIE BD1 HEAT EXCHANGER [OEPRESSURIZATION~~

~~STAGE 4~~

~~(SILVERED STAGE 3 MOLECULAR SIEVE~~

~~STAGE 2~~

~~STAGE 1~~

~~STAINLESS DEXTER PAPER GRID STEEL MOLECULAR SIEVE FILTER~~

~~FIG 4 - COMEDIE BD1 SLOWDOWN FILTERS FIG 5 - DEPRESSURIZATION SECTION FILTER 2.2. Out-of pile section~~

~~The role of this section is to set up the flow rate and the chemical composition of the coolant gas. It consists of:~~

~~fule lth flo w• filter which facilitate operatioe sth n under clean condition trappiny sb g solid particles ("dust" condensibld )an . eFP~~

~~blower e circulators th ga r • ,o , whic capablhs i achievinf eo mase gth s flowo t rat p eu 70 g.S'1 helium under 60 bar pressure~~

~~• facilitie analysiss ga r purificatios sfo ga , injectiod nan desiref no d chemical impurities.~~

~~3. Irradiatio depressurizatiod nan n phases~~

~~3.1. Steady state irradiation:~~

~~This phase is to establish and measure steady state operating conditions prior to initiatin transiene gth blowdowr o t n phasexperimente th f eo .~~

conductes Itwa cycle3 r SILOdayse d3 fo th s(6 n i )E reactor under clean conditions. The operating conditions were maintained effectively constant throughou irradiatioe th t n except to the coolant impurities which varied somewhat.

~~These parameters are:~~

~~bar0 6 s coolan pressur s : tga e coolan temperaturs tga : e [inlet 650° C fuel element : < [outlet 720° C~~

~~[inlet 720° C heat exchanger : < [outlet 295° C~~

~~Chemical composition of the coolant gas (typical):~~

~~H20 2.3m pp +1 m pp 3 ± 2 1 2 H Co 6+1 ppm~~

~~During the steady state irradiation, the release of noble gases (Xe and Kr) has been measured continuously on line.~~

~~The noble gas release was first very low :~~

~~cyclt f predictioo 1s e% e 2 th - r nfo - a rapid increase at the beginning of the 2nd cycle - the predicted values were reached at the beginning of the 3rd cycle.~~

This release profile particle P showDT thas e su s th t have starte faido t l significantly cycled beginnine 3r th t e . a cycld Thith slowea durind f 2n gs o si e an e rgth rate than i previous tests particle PETTEwitP R hDT HF HFId n sN (i ORNL)t an R a .

~~The durin 1 releas 13 steade I g th f eo y state irradiatio deposited nan whole th dn ei loop is about 650 mCi.~~

~~150 3.2. Depressurization phases (fig: ) .6~~

~~These phases called blowdown tests were devoted to the lift off of the FP deposited principally on heat exchanger walls, to transport and to trap them on specifically designed filters.~~

~~One of the three bundles of the exchanger was isolated prior to the blowdowns and was considere referenca s da e (i.e. plateoue th , f steady-stato t d distributioen e th et na irradiation phase).~~

~~4 blowdowns of increasing levels have been carried out in a very short time after the irradiatioe th f o d nen phas ordeen i keeo rt maximupa mf shoro t half . lifeFP~~

~~For turbulent flow insid circulaea r duct (i.e heae .th t exchanger tubes), strengtf ho blowdown sheagivee s i th y nrb ratio (SR: )~~

~~S ,0.75 L75 -0.58 **]n PN;~~

with P pressure of the coolant gas V flow velocity at the outlet of heat exchanger TG gas temperature at the outlet of heat exchanger B value under blowdown condition N value under steady state condition

The shear ratio is defined to be the ratio of the transient wall shear stress to the wall shear stress during steady-state irradiation. It is used to determine the amount of radioactive plate out that is lifted off from the circuit surfaces and entrained in the coolant circuit.

~~COMEOIE 2.00 + BS.OO X~~

~~LOOP PRESSUREX bar~~

~~10000 * SO X~~

~~BLOWER SPEED mm~~

~~MASS FLOW RATEX~~

~~FI BLOWDOW- G6 N SEQUENC7 1. = R ES~~

~~151 4 blowdowns corresponding to :~~

SR = 0.72 1.7 2.8 5.6 have been performed, mainl flovaryins y yb wga velocite gth raisiny yb blowee gth r speed d maintaininan constana t a gt i t leveminutes2 r fo l rapiA . d depressurizatioo t r n froba m0 6 6 bar has taken place just after this short holding period.

~~4. Plateout and reentrainment of FP~~

~~Post irradiation examination has been performed by y and p spectrometries to measur distributioamoune e eth th d an t f radionuclideno s withi varioue nth s componentf so the loop.~~

~~4.1. Heat exchanger:~~

~~The figures show the FP profiles on bundle W (reference, isolated prior to blowdowns) and on bundles U and V (subjected to 4 blowdowns).~~

preferentialls i ) 7 (fig1 . 13 I y deposite coldese th dn i % tbundlee 90 zon th > n t eo sbu of this nuclide passed throug heae hth t exchanger. take w ef I compar into . t 479: i oW s i r 5uC V fo 532 o eactivitt i 1 + 2U 13 uC f I yo consideration global I 131 activity (650 mCi) within the loop we can say that:

f 113• o 1onl % yhav2 e been deposite heae th tn dexchangeo r • during blowdowns f iodino % ebundleV 5 deposite5 ,d bees an sha nU moved.an do sa result of the chemical desorption. Howewer, this is < 1 % of the total I 131 plateout in loope th .

~~140%~~

~~8 120% § £ £ioo% i 5 0> 80%~~

~~Junction 8 60%~~

~~.1 «* + S « K 20%~~

~~0% ——(- 1300 Inlet 1800 2300 28OO 3300 3800 outlet Axial position (mm)~~

~~7.113G FI 1 DISTRIBUTION PROFILE BUNDLER SFO W , V , SU~~

~~152 AgJMOmJfig.S)~~

maie Th n110g A parconcentrates i mf o t bundlee inlee th f th o t n di s (the hottest zone). The heat exchanger has been very efficient in trapping that FP. The amount of Ag 110m is equivalent in the three bundles (nearly 380 uCi on each of them). Slowdowns had no influence, neither on the deposited amount nor on the profile of Ag 110m located inside the bundles.

~~Cs 134 and 137 (fig. 9 -10):~~

Those two nuclides deposit the same way in the inlet zone of the tubes, profiles and amounts being nearl thresame e yth th een i bundles 110msamg e A Th .s e,a total activities 134s C .i uC 9 5 d an 7 eacn i 13 s h C bundl i uC vere e8 ar y 47 similar o t 1 ,44

~~Slowdown noticeabla o t sd le hav t e eno modificatio t thosno e nuclides depositen di bundlesV d an .U~~

~~) (fig2 T11 .e13~~

Deposition profiles of that nuclide are the same for the three bundles. It is located in the hot zone of the bundle for the main part. The general behaviour is the same as the one of Ag 110m, including the effect of blowdowns.

~~Total activitie eacsn i h bundl bundlr alsfo e e2 , ar o e W 13 ver e yT i similarmC 6 :8, 8mCiTe132forUandV.~~

~~4.2. Depressurization section~~

~~1131 (fig. 12)~~

havinbundles1 V 13 d 75 I g an 6f o migrate ,uCiU havf % ,o 3 t thae1 d ou s beei t n trapped on the four filters, 60 % of which (451 MCi) being found of the filter n° 2 (SR = 0.72). Howewer, this represents < 1 % of the total 1131 plateout in the loop. The activity (451 uCi) can't be representative of the first blowdown (SR = 0.72) but, more probably, is linked to Iodine desorption phenomena, during the rather long time (5.8 h) when the loop was brought back to operating temperatures with the electrical heaters, prior to the first blowdown.

~~Ag 110m ) (fig13 .~~

The trappe filters4 d e activitieth ) n agreementhan i o ver (2,3i e s w ti s ar ylo uC 8 t with what have been seen when examining the bundles. We find 80 % of the total Ag 110m trapped on the filter corresponding to the highest shear ratio (SR = 5,6).

~~-15)(fig4 7 1 . 13 : d an 4 Cs13~~

~~collectee Th d activitie: ver e w sar y lo Cs137 : 1.36uCi Cs134 : 0.156 uCi blowdown6 5. = f measureR o S linke% e .e 0 ar th 7 s do t d C~~

~~Te132~~

~~About 1.2 uCi has been trapped on the four filters.~~

~~153 120% |Ag - 110m J . BUNDLEU 120% BUNDL. E U " BUNDLE V 100% tf> " BUNDLE V ' BUNDLEw 100% ' BUNDLE W 80% 80% 800H 0)1 60% S • 60%~~

~~40% d) 40% * * * 8 " o u n i t " I 20% 20%~~

" . n I « • * 0% i i i i i i 0- % —t ——|—— i - < i • — —I i 1300 Inlet isoo 2300 2800 3300 3aoo outlet 1300 Inlet isc 2300 2800 3300 3800 Outlet Axial position (mm) Axial position (mm)

~~110g A . m8 DISTRIBUTIOG FI N PROFILE BUNDLER SFO W , V , SU DISTRIBUTIO7 13 s C . 9 G FI N PROFILE BUNDLER SFO W , V , SU~~

~~— 1»°* . BUNDLE U a BUNDLE V 600 J5 140% |C»134 ' BUNOLC W m ? 120% o 500 - 100%~~

~~3 80% 400 £ eo%~~

~~j! 40% g~~

~~two Inlet 1800 2800 3300 3BOO Outlet 1300 Inlet i8oo Axial position (mm) 3300 jaoo outlet Axial position (mm)~~

~~2 DISTRIBUTIO13 e T . 11 TEMPERATURD G FI NAN E PROFILER SFO FIG 10. Cs 134 DISTRIBUTION PROFILES FOR BUNDLES U, V, W BUNDLEW , V , SU I »•«*"» imi — 1 ^ 4,Hft01~~

~~fi 4.00ft 01 t - 03g 0,Wi JtOf tOI • 300C tOl £ I.OOftOO~~

~~JKX.OI f.MEtOO LOW 1 01~~

~~I.HftOl " 1.00*1 oo~~

~~I.OOftOl I.OOf-01 1,00ft 01 o.oot too FHTER?(SR*072) FILTERS (SR = 1 7) FILTER 4 (SR = 2 8) FILTER 1(SR = 5 8) FILTER 2 (SR » 0 72) FILTER 3 (SR = 1 7) FILTER 4 (SR = 2 8) FILTER 1(SR • 5 8)~~

~~FIG 12. DEPRESSURIZATION SECTION: TRAPPED 1131 ACTIVITY FIG 13. DEPRESSURIZATION SECTION : TRAPPED Ag 110m ACTIVITY IN IN FILTERS 1 to 4 FILTERS 1 to 4~~

~~Q t 201-01 g l.JMtOO |Ct 137 _j (Cs134 ~| " S 1 IOOC-01 • 1 I.OOftOO~~

~~e oof 42 •,00f-0l •~~

~~tOC* 41 a.ooi-01 •~~

~~4.00C-01 4.001-01~~

lOOf-02 l.OOf-OI . • • . O.OOftOO ---«•-••- , -I O.OM.OO FILTE (5RR2 » 072) FILTE (SRR3 ) 7 *1 FILTEfM(SR = 28) ) 56 = FILTE R RHS FILTER 2 (SR* 072) FITER3(S) 7 1 R= F\TIR4(SR°28) FIT{H1(SR»6«)

~~FIG 14. DEPRESSURIZATION SECTION : TRAPPED Cs 134 ACTIVITY IN FIG 15. DEPRESSURIZATION SECTION : TRAPPED Cs 137 ACTIVITY IN Ul FILTERS 1 to 4 FILTERS 1 to 4 5. Conclusions~~

~~resulte Th s show that:~~

~~• deposition profiles are dependant on the chemical nature of fission products~~

• an important redistribution of I 131 (55 % of the initial heat exchanger bundles U and V inventories but < 1 % of the total I 131 plateout in the loop) occured during the thermal conditionin looe th p f g o befor firse eth t blowdown. wele ar l 2 fixe13 de withiT d depositioe nan th 7 110mg A 13 134 s s nC , • C ,sectio d nan blowdowns did not move significant quantities of them.

extrapolationy b • fractionae th , l liftoff durin rapiga d depressurizatio MHTGn a f no R l radionuclidesal r fo % shoul1 < ,e dincludinb g isotopesI , sinc peae eth k shear rati< os i blowdowe th 1.1d an . completns i minutes w fe a en i .

~~REFERENCES~~

~~[1] MEDWID W. , Specification for COMEDIE Test BD1 - DOE HTGR - 87095, Rev. H, November 92 General Atomics, San Diego, Ca.~~

~~156 LICENSING EXPERIENC HTR-1E TH F 0EO TEST REACTOR~~

~~Y. SUNU X . ,Y Institute of Nuclear Energy Technology, XA9642787 Tsinghua University, Beijing, China~~

~~Abstract~~

A 10MW high temperature gas-cooled test reactor (HTR-10 beinw no g s i ) projecte e Institutth y b f dNucleao e r Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authorit f 199yo d aroun4 en afte e perioda rth f aboud o yea e f safeton to r y review of the reactor design.

HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpos thif eo s test reactor projec demonstrattesd o t an s ti t technologe eth safetd yan y feature advancee th f o s d modular high temperature reactor design reactoe Th . r uses spherical fuel elements with coated fuel particles reactoe Th . e steath r munid an generatot r unie ar t arranged in a "side-by-side" way. Maximum fuel temperature under the accident condition of completa e los coolanf so limites i t valueo dt s much lower tha safete fuee nth th lr y fo limi t se t element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design reactoe th , alss ri o calle Tese dth t Module.

HTR-10 represents among others als licensinoa firse th gte sides i challenge on t i , e th n O . helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review.

In this presentation e licensinth , g criteri f HTR-1o a e discussed0ar e organizatioTh . d nan activitie safete th f yso revieconstructioe th r wfo n permit licensin describede ar g . Some th f eo main safety issuelicensine th n si g procedur addressede ear . Among these are r examplefo , , fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of great reference value for the modular reactor concept.

~~1 Introduction and Background~~

Presently, a 10MW high temperature helium cooled test reactor (termed HTR-10) is being projecte Institute th y db Nuclea f eo r Energy Technology (INET Tsinghuf )o a Universitye Th . reactor wil erectee site b lINEf th eo n do T whic abous hi t norte 40kth Beijinf mho o t g city. The HTR-10 test reactor is a major project in China's High Technology Programme.

The HTR-10 test reactor uses spherical fuel elements which are made completely of ceramic materials. Uranium dioxid s nucleaa e e for th f coate mo rn i fue s i dl particles whice ar h disperse graphite th n di e matrifuee th l f elementsxo . Graphite serve s neutroa s n moderator

157 and reflector as well as the main structural material of the reactor core, so that the reactor has a practically full-ceramic core a largwhic s eha h heat capacit s high-temperaturei d an y - resistan s coolanA t t serve e iners heliuth s ga t m which causes practicall o corrosion y n problems and plays no part in the reactor neutron balance

FoHTR-1e th r 0 test reactor, decay heat remova activy an en o lcoolin l doe t al rel t sno ya g systems At a complete loss of coolant accident, the maximum fuel temperature remains under the limit valu emargi g witbi ha n Reactor shut down system e placear s de sid onlth en yi reflector No control rods would have to be inserted into the pebble bed so that damages to the fuel elements are avoided The reactor unit and the steam generator unit are arranged in a so-called "side-by-side" way, so that the reactor and the steam generator can be easily isolated from each other under accident condition orden si proteco rt ceramie th t creactoe corth f eo r anmetallie dth c structur steae th mf eo generator These design feature HTR-1e th f o s 0 test reactor represent the advanced modular design of high temperature gas cooled reactors (HTGR)

In terms of reactor types, HTR-10 is the first of its kind to be built in China Besides, the reactor design incorporates many advanced feature directioe th n i s f passivno e safety From these point f viewo s HTR-1e th , 0 reactor represent licensing bi sa g challengee botth o ht sidee on applicane ,regulato e th ther th o t n ed O texis an t renoug no t h standards, coded san guide Chinn si a directly applicabl s coolega a othee o et dth w reactorn rno o side s i d t i ,An s world-wide problem for regulators how to properly treat passive features in the advanced reactor designs In the next sections, the settlement of the main safety issues in the licensing procedure wil addressee b l d

Folicensine th rconstructioe th f go n permit HTR-1(CPe th f )o 0 test reactor licensine th , g authorit Nationae th s yi l Nuclear Safety Administration (NNSA) whic technicalls hi y backed up by Suzhou Nuclear Safety Center and Beijing Nuclear Safety Center The applicant is the Institut Nucleaf eo r Energy Technolog Tsinghuf yo a University

~~2 Regulatory Criteria~~

As state therw dno e above o exist Chinp n i u t , t enougano h nuclear standards, coded an s guides specifically compile r higfo dh temperatur s coolega e d reactors whic e directlar h y applicable to the HTR-10 reactor Before the CP licensing started, two documents had been compiled under the organization of NNSA Design Criteria for the 10MW High Temperature Gas-cooled Test Reactor Standard an ' d Conten Formad Safete an tth f o yt Analysis Reporf o t 11 the 10MW High Temperature Gas-cooled Test Reactor , which were supposed to serve as 121 the licensing basis of the test reactor In the actual licensing procedure, stronger reference is secone madth o et d documen tfirse e thath on t o nt

From the nuclear point of view, the HTR-10 reactor is much smaller than nuclear power plant reactors sinc thermas eit l ratin onls gi y 10M maie WTh n purpos reactoe th f eo r projeco t s i t test and prove its main technical and safety features rather than to provide commercial power But HTR-10 has not been regarded as a research reactor because the overall purpose of the reacto teso t s ti r power generation technology abovBasee th n deo considerations, following principal guidelines have been followed in the licensing procedure concerning what standards and/or codes are taken as licensing basis or references

158 • As top level regulatory criteria, the nuclear safety related standards, codes, regulatory guides issue NNSy db A (HAF series), which covers site selectio f nucleano r power plants (NPP), design of NPP, quality assurance of NPP, etc., should serve as fundamental basis for licensing criteria. • Reference made sar internationao et foreigd an l n standards, code guidesd A san US ,f o e.g. (RG series, ASME, General Design Criteria of High Temperature Gas Cooled Reactors), of Germany (KTA Regeln) and of France (RCC series). • HTR-10 deserves special treatment in specific cases considering its lower power rating, its natur tesa s tea reacto especialld ran designes yit d inherent safety * features.

~~Licensin3 g Activit Procedurd yan e~~

~~3.1 Pre-application activities~~

Since the HTR-10 test reactor is the first high temperature helium cooled reactor to be built in China, ther e lesar es experiences both fro e regulatomth r sid licensinn i e d froe an gmth applicant side in design. Therefore, there had been many interactions between the licensing experts and the design engineers in the form of e.g. seminars given by the design engineers to introduce the HTR-10 design and general features of high temperature gas cooled reactors to licensine th g experts. These communications helpe involvee dth d personne exchango t l e ideas and opinions on an early stage.

As mentioned above, NNS organized Aha o establisdt h technical documents l1'21 before th e licensing procedure started secone Th . d document, namel Standare yth d Conten Formad an t t of the Safety Analysis Report of the 10MW High Temperature Gas-cooled Test Reactor, which defines the content framework of the Preliminary Safety Analysis Report of the 10MW High Temperature Gas-cooled Test Reactor (PSAR)I3), has guided the compilation of the document.

Followin e principlgth f "earlieo e r involvement", some licensing stafbeed ha fn involven di some pre-application wor e sitth ke n sucselectioi s ha n activitie assuro t s emora e smooth licensing procedure.

Before the licensing procedure started, the applicant had got the HTR-10 project approval from the State Education Commission and the approval of the Feasibility Study Report of HTR-10 from the State Science and Technology Commission. The "Environmental Impact Report of the 10MW High Temperature Gas-cooled Test Reactor" had been approved by the State Environmental Protection Administration.

~~3.2 Licensing activities and procedure~~

~~The application for the Construction Permit of HTR-10 was submitted to NNSA in December 1993 with the attached documents, of which the PSAR[3) and the Quality Assurance~~

~~Programm e 10Mth W f o e High Temperature Gas-cooled Test Reactor (for Desigd an n~~

1 Construction Periods e mai)th (QAP)ne ar technica 14 l document e revieweb e o th t s y b d licensing personnel. From then on the licensing procedure started formally and lasted for one year. The applicant, namely the Institute of Nuclear Energy Technology, got the Construction Permi Decemben i t r 1994.

~~159 The licensing procedure goes with the following main activities undertaken:~~

• The reviewers (licensing personnel) raise technical questions in written form. Altogether, more thaquestion0 n70 s wer severan ei raiseP PSAln QA dbatcheso d Ran . raisee Th d question• answeree sar HTR-1 e th y db 0 design engineer written si n form. • Meetings are held between reviewers and design engineers to discuss and address technical question issuesd an s . During these meetings, clearance r solutiono s o somt s e question r problemso founde sar r agreemento , s upon some technical issue reachede sar . Unresolved issue e documentear s fore th f so-callem o n i d d work-sheets. These work- sheets are worked on further by the designers for further interactions. • Topical meeting hele sar d between reviewer desigd san n engineer somn so e special issues in site selection, digital protection system design, quality assurance. • The Nuclear Safety Expert Committee is consulted about some special issues and about the overall evaluation of the licensing personnel of the HTR-10 test reactor on the CP application stage. • The licensing authority finally reaches a favorable Safety Evaluation Report which leads to the issuing of the construction permit of the HTR-10 test reactor

~~4 Main Licensing Safety Issues~~

~~4.1 Fuel elements~~

The designed passive safety features of the HTR-10 test reactor are fundamentally based on excellene th t fission product retaining capabilit fuee th reasonable l f elementsth yo l al r yFo . postulated accidents, both withi desige nth n basi beyond san d that, radioactive nuclidee ar s retaine e fueth l n i delement s well enoug thao s h t unallowable radioactive release inte oth environment will not take place. Therefore, it has been a core issue during the licensing to make sure whether the fuel elements used for the HTR-10 reactor will really behave as good as they are designed for. It is planned that sample fuel elements are to be irradiated to designated conditions covering burn-up, fast neutron dose and irradiation temperature before the fuel element operatee sar reae th l n reactodi thoso rt e conditions oxidation A . e n th tes f o t fuel elements under severe accident conditions is also planned to be made.

~~4.2 Source term~~

A mechanistic methodolog s adopteyi r determinindfo radioactive gth e source term. Severe core damaget arbitrarilno e ar s y postulated s doni r largt i fo e s ea , water cooled power reactors, where large amoun radioactivitf o t y would postulatee havb o e t release e b o dt t dou fuef o l element severo t e sedu core damages whicd an , h woul drequiremene th lea o dt a f o t pressure-containin leak-tighd gan t containment design.

release Th f radioactiviteo calculates yi d specificall thosr yfo e individual demanding accidents which lead to the most release of radioactive nuclides from the fuel elements. The release of radio-nuclides from the fuel kernel through the surrounding coatings and the matrix graphite to the helium coolant is calculated on a diffusion-sorption basis, where the selection of calculation parameters is based mostly on the German experimental results and literature. Where necessary, conservative factors are put into the analysis.

~~160 Acciden3 4. t analysis~~

~~As usual, design basis accidents (DBA classifiee ar ) severan di l categories These ear~~

• increase of heat removal capacity in primary circuit • decrease of heat removal capacity in primary circuit • decreas primarf eo y flow rate • abnormality of reactivity and power distribution • ruptur primarf eo y pressure boundary tubes • anticipated transients without scram (ATWS)

The reacto designes i r d against these accident analysie Th s f thesso e accident dons si e with conservatis analysie mTh s results show excellent safe respons reactoe th f eo accidentao t r l events Withi framewore nth f DBA o kaccideno n , t would lea relevano dt t releas f fissioeo n products from the fuel elements

Great attention and effort has been given to the treatment of severe accidents A number of highly-hypothetical postulated accident selectee s ar analyze e b do t d These hypothetical events mainly include

• long-term failure of the reactor cavity cooling system • simultaneous withdrawal of all control rods at power operation and at reactor start-up • failure of the helium circulator shut-down • simultaneous ruptur steal al f meo generator tubes • ruptur crose th sf eo duc t vessel

In selecting these severe accidents, reference is made to the licensing experience of MHTGR in USAHTR-Moduf o d [5an ] Germann i l analysee Th y f severso e accidents show that under these highly-hypothetical circumstances, severe damage fuee th l o elementt s s e woulb t dno expected which would lead to impermitted release of radioactivity into the environment

~~Safet4 4. y classification~~

Because the HTR-10 test reactor is designed on the inherent safety philosophy, safety classification f systemso componentd san s departur dons i water t i efo y er wa frocoole e mth d power reactors For example, primary pressure boundary is defined to the first isolation valve Steam generator tubes are classified as Class II component Diesel generators are not required highls a e b y o qualifiet thoss da e use largr dfo e water cooled power reactors, sinc systemo en s or components with large power demand would requir immediatn a e e diese th star f lo t engines at a plant black-out accident

~~4.5 Containment design~~

Based on the characteristics of inherent safety of the HTR-10 test reactor, no pressure- containin leak-tighd gan t containmen designes i t concrete Th d e compartments, which houses the reactor and the steam generator as well as other parts of the primary pressure boundary and whic preferabls hi y calle s confinementda , together wit accidene hth t ventilation system, serve as the last barrier to the radioactivity release into the environment During normal operation confinemene th , ventilates i t kepe b o tdt sub-atmospheric Whe integrite nth e th f yo

161 primary pressure boundary is lost, the primary helium coolant is allowed to be released into the environment without filtering because of its low radioactivity content Afterwards the confinement is ventilated again, gases in it will be filtered before they reach the environment

~~Summar5 y~~

The licensing activitie constructioe th f so n permiHTR-1e th f o t 0 test reacto overale rar l well organized in a rather tight time-framework The evaluation of the licensing body on the safety favorites the reactor safety design and has led to the issuing of the construction permit of the HTR-10 test reactor

Experiences in licensing HTR-10 are of great reference value for the modular concept of high temperature gas cooled reactors. The main safety issues would be roughly the same with the modular concepmethodologe th d an t y use licensinn di HTR-1e gth 0 shoul greao t e db t extent applicable when licensin gmodulaa r power reactor

~~REFERENCES~~

Institut f Nucleao e r Energy Technology, Tsinghua University Design Criterie th r fo a 10MW High Temperature Gas-cooled Test Reactor, 1993 Institute of Nuclear Energy Technology, Tsinghua University. Standard Content and Forma Safete th f yo t Analysis Repore 10Mth f Wo t High Temperature Gas-cooled Test Reactor, 1993 Institute of Nuclear Energy Technology, Tsinghua University Preliminary Safety Analysis Repor 10Me th f Wo t High Temperature Gas-cooled Test Reactor, 1993 Institut f Nucleao e r Energy Technology, Tsinghua University Quality Assurance Programme of the 10MW High Temperature Gas-cooled Test Reactor (for Design and Construction Periods), 1993 Williams . KingL N WilsoT , J. , n Draft Preapplication Safety Evaluation Repore th r fo t Modular High-Temperature Gas-Cooled Reactor NUREG-1338 March 1989

~~162 GAS-TURBINE POWER CONVERSION SYSTEM DEVELOPMENT~~

~~Session4 DEVELOPMENT OF COMPACT HEAT EXCHANGER WITH DIFFUSION WELDING K. KUNITOMI, T. TAKEDA lllllllllllllimilll»wmilll«i«ii Department of HTTR Project, Japan Atomic Research Institute, Ibaraki~~

~~T. HORIE, K. IWATA Heat Exchangers Division, Sumitomo Precision Products Co., Ltd., Hyogo~~

~~Japan~~

~~Abstract~~

A plate fin type compact heat exchanger (PFCHX) normally uses brazing for connecting plates fins.and However, reliabilitythe brazingof insufficientis when PFCHXs usedlonga are for duration primaryas secondaryor components nuclearin plants. Particularly, PFCHXs used as a recuperator of gas-turbine plant with the High Temperature Gas-cooled Reactor (HTGR) or Intermediate Heat Exchanger (IHX) in future generation HTGRs need high reliability in a high temperature region. We have been developing PFCHXthe with diffusion welding between platesand fins. The tensile and creep strength in the diffusion welds are superior to those in the brazing especially in high temperature condition. The developing PFCHX consisting of Ni based HastelloyXR plates expectedis usedbe to °C. over900 Prior to the development of a full scale PFCHX, the small PFCHXs with the diffusion welding were designed, manufactured and installed in a test loop to investigate the welding strength and reliability. The early tests showed that reliability of the diffusion welding is very high, and the PFCHX with the diffusion has a possibility for the IHX or recuperator. Thermal performance tests were also carried obtainto out effective thermal conductance. This paper describes design the test and results smallthe of compact heat exchanger with diffusionthe welding.

1. Introduction The High Temperature engineering Test Reactor (HTTR) [1] being constructed in Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium- cooled reactor with an outlet gas temperature of 950°C. The HTTR will be used to establis upgradd han technologicae eth l basi r advancesfo d HTGR conduco t d san t various irradiation test innovativr sfo e high temperature basic researches alse b on ca t I . use r demonstratiodfo f variouno s nuclear process heat applications thastrongle ar t y required from various standpoint 2 emissios sucCO s ha n reduction, efficienf o e us t . on energo s d yan The Intermediate Heat Exchanger (IHX) [2] which can be used over 900°C condition essentian ia s l componen r procesfo t s heat applications Helicalle Th . y coileX dIH (HCIHX bees ha nf 10M )o develope W installed dan primara dn i y cooling system (PCS) HTTRe inth . Variou test desigD d ssan R& n works were carrie develoo t t dou p materials and component structure HCIHXe th HCIHe n si Th . X wil testee b l higdn i h temperature

165 condition during the HTTP operations in order to investigate thermal-hydraulic characteristic structurad san l integrity technologe Th . yHCIHe basith r Xsfo wile b l established in these tests in the HTTR. The problems of the HCIHXs are their size and cost. In the HCIHX used in a 400-600 MW thermal HTGR plant, its manifold to support heat transfer tubes cannot be easily manufactured becaus weldinea g structur t propeno es i keeo rt p creep straid nan creep fatigue damage lower than an allowable level. Even if a big forging manifold is manufactured plane th , t loses economical competitiveness othee th rn O hand. a , compact heat exchanger (CHX) has 20-30 times larger heat transfer area per volume than the HCIHX. The CHX is expected to be the best heat exchanger in the future generation HTGRs with process heat application. It can be also used as a recuperator iGas-Turbinna e wit Modulae hth r High Temperature Reactor (MHTGR). In our study, a new CHX can be used over 900°C condition is designed and manufactured. The CHX consists of only plates welded with diffusion. The tensile and creep strengths in the diffusion welding are superior to those in the brazing especially in high temperature condition. Major feature HCIH e HTTe describeth e th presend f ar so Xn Ri an X CH tn di Chapter 2 and 3, respectively. The design characteristics of the new CHX with diffusion welding and its test and results are described in Chapter 4.

2. Intermediate Heat Exchanger for the HTTR Figure 1 shows a reactor cooling system of the HTTR. The reactor cooling system HTTe oth f R consist secondarPCSe a sth , y helium cooling system, pressurized water cooling syste vessed man l coolin consistgS system maiPC a f e so n Th .coolin g system, an auxiliary cooling system. The main cooling system has two heat exchangers, the 10MW HCIHX and a 30MW pressurized water cooler. The HCIHX for the HTTR is a vertical helically-coiled counter flow type heat exchange whicn ri primare hth y heliu shel e flow s secondarth me ga n l th s o sid d ean yn i the tube side as shown in Fig.2. The major specification is shown in Table 1. The primary helium gas of the maximum 950°C enters the HCIHX through the inner tube in the primary concentric hot gas duct. It is deflected under a hot header and discharged around the heat transfer tubes to transfer primary heat to the secondary cooling system. It flows to the primary circulator via an upper outlet nozzle and returns between the inner and outer shell in order to cool the outer shell.

On the other hand, the secondary helium gas flows downwards inside the heat transfer heates tubei 905°Co d t p san d u flowt I . s upwards insid s centrae eth ga t ho l ductinnee Th r. insulatio installens i d insid innee eth r shel maintaio t l temperature nth e of the inner shell under the allowable one. The thermal insulator outside and inside the central hot gas duct prevents the heat transfer between the primary and secondary coolant except the heat transfer area on the heat transfer tubes so that high heat transfer efficiency can be obtained, and the temperature of the central hot gas duct is maintained under the allowable one. The pressure of the secondary helium gas is controlled higher than that of primary helium gas for prevention of fission product release even if the heat transfer tube shoul brokene db . A tube support assemblies hol heae dth t tubes. Both ducs centrad ga an tt ho l heat tube support assemblie hangee sar d wit vessethaha o s t p thermato l l expansion cannot be constrained. Material of the heat transfer tubes and the hot header is Hastelloy XR, and inner outed an r shell mads i 1/4Cr-1Mo2 f eo Ni-base Th . e Cr-Mo-Fe superalloy Hastelloy developeo s s X Rhavo wa t s superiodea a r corrosion resistance unde exposure rth eo t

~~166 CONTAINMENT VESSEL VCS~~

~~WATER PUMP AIR COOLER L~~

~~AIR COOLER~~

~~ACS MCS~~

MC Mai: S n cooling system Intermediat: IHX e heat exchanger PPW CPrimar: y pressurized water cooler PGC : Primary gas circulator SPWC : Secondary pressurized water cooler SG Secondar: C circulatos yga r ACS : Auxiliary cooling system AH Auxiliar: X y heat exchanger AGC : Auxiliary gas circulator VC Vesse: S l cooling system Concret: CS e shield

~~Figur e1 Reactor Cooling Syste HTTe th mf Ro~~

~~167 Secondory Helium (to Secondory PWC) Secondory Helium Urom Secondory PWC)~~

~~Cold Header~~

~~Primory Helium Tube Support Bsam o Primord s Co y Circulator )~~

~~Primory Helium (from Primar s CirculolorGo y ) Inner Shell Ouler Shell Tube Support Assembly Cenlrol Hot Gcs Dud (Center Pipe) Thermal Insulator Helically Coiled Heal Transfer Tube Hot Header~~

~~Primory Helium ho Reader) Primory Helium Hrom Reoctoil~~

~~Figure 2 Bird's eye view of the IHX in HTTP~~

~~the HTTP primary coolant. The 2 1/4Cr-1Mo is also superior in anti-corrosion and strengt highn i h temperature condition.~~

3. Compact Heat exchanger The CHXs widely used in conventional plants such as fuel cells power generation plant, co-generation plant and gas-turbine plant have several advantages described as follows; largA ) e(1 heat transfer are hear ape t exchanger volumofferede b n eca . (2) Variou stype n kinflofi d f wdsan o pattern selectede b n sca . (3) Compact sizhigd ehan thermal performanc obtainede b n eca .

~~168 Tablel Major specificatioX IH f no~~

Type Helically-coiled counter flow Design pressure Outer shell 4.7 MPa Heat transfer tube etc. 0.29 MPa Design Temperature Outer shell 430°C Heat transfer tube 955°C Operating condition Rated High temp, operation operation Flow rate of primary He 15t/h 12t/h Inlet temp, of primary He C ° 0 85 950 °C Outlet temp, of primary He 390°C 390°C Flow rate of secondary He 14t/h 12t/h Inlet temp, of secondary He 300 °C 300°C Outlet temp, of secondary He gas 775 °C 905°C Heat capacity W 10M Dimension Diamete heaf ro t transfer tube 31.8mmOD Thicknes f heaso t transfer tube 3.5mm Outer diamete outef ro r shell 2.0m Total height 10.0m Material Outer and inner shell 21/4Cr-1 Mo Heat transfer tube Hastelloy XR

Figure 3 shows a typical CHX used in conventional plants [3]. It consists of a plate find ssan core, heade nozzlesd an rplatese Th . , finsblazind an , g metalse ar , bonded by brazing in a vacuum furnace. Brazing technology for the CHX was well developed strengtd ,an anti-corrosiod han n characteristic brazine th f so g material were confirme severay db l tests. Their composing material stainlesse sar , aluminuo s d man on. These CHXs have never use primars da y heat exchanger unclean si r power plants. However, they hav possibilitea usee b do yt less than 600°C.

. Developmen4 f Compaco t t Heat Exchanger with diffusion welding 4.1 Development items and schedule A final goa f thi o luse e primarsa b studs tha dn X develoa o t ca tys CH i e pth intermediate heat exchanger in a future generation HTGR plant or recuperator in the MHTGR with the direct gas-turbine (GT-MHTGR).

~~169 SEPARATE PLATE~~

~~CORRUGATED FIN~~

~~V " ^SPACER BA R~~

~~BRAZING FILLER METAL~~

~~Figure 3 Typical CHX used in conventional plants~~

keeo t Requiremente p ar primar X IH e y th coolan o st t during normal operations and off-normal conditions and transfer heat efficiently. The boundary between primary and secondary coolant must be assured when a significant system failure occurs in a secondary system. The recuperator in the GT-MHTGR is not used in higher temperature condition than the IHX. However, it is not easy to meet their requirements because the recuperator is a key component to improve efficiency for the GT-MHTGR[4]. Requirement recuperatoe th o st gas-turbine th n ri e systefollowss a e mar , (1) Thermal effectiveness is over 90%. ) (2 Pressure dro lesps i stotaf o tha l% npressur2 e drop. (3) Pressure difference between low and high temperature helium gas is the maximum 5MPa. (4) Life yearstim0 4 es i . (5) Structural integrit assureys i off-normadn i l transients, start-u shut-downd pan . (6) In Service Inspection is necessary when it is used in a direct cycle plant. Effectiveness, pressure dro differentiad pan l pressure betwee e sideo th nr tw sfo recuperator used in an existing gas-turbine plant are approximately 88%, 6% of GT plant total pressure drop and 1.4MPa, respectively. In addition, components with the brazing welding have never been used in primary components. The CHX with diffusion welding bese th tf o candidate e ison solvo st e above problems. In the first stage, the feasibility study of the CHX with the diffusion welding will be carried out by designing and manufacturing a small CHX. In addition, small specimens diffusioe ofth n welding wil manufacturee b l measurdo t e their strengt observd han e their welding surfaces. In the second stage, after comparison between the CHX with the conventional brazing and new diffusion welding is conducted, the best CHXs for the GT- selectede finaar e th X MHTG l n I IH stage . d fule Ran th l, scale model test wile b l conducte df necessaryi .

170 4.2 Design characteristic witX h CH diffusio e th f so n welding In order to establish technological base for the CHX with the diffusion welding, the small CHXs wit diffusioe hth n welding were designed, manufacture installed dan d in an air loop to investigate the welding reliability and confirm the structural integrity. Tabl eshow2 s major design characteristic smale th f slo CHXs with diffusion welding. They only consist of Hastelloy XR plates with flow paths. The flow paths are machined b numericaya l control machining system. Those plate weldee sar d wit diffusione hth . separate th e us et Thefinno so yshowd n Fig . Thes.3 unique ear e characteristice th f so new CHX and it is easy to be manufactured comparing with the CHX with brazing diffusion. Moreover, considering thauses i t i td over 900°C condition, Hastellos i R yX selecte platee th r sdfo material. Hastellosame th es i materiaR yX thas a l t usee th dn i HCIH HTTe superioth d XRn i an creen i r tensil d pan e strengt higa n hhi temperature region. Figure 4 shows the cross sectional view of the CHX.

4.3 Diffusion welding metho resultd dan s First, the best condition for the diffusion welding is determined by the tests using two small thin plates. Thos chambea n i e t pressurizeplated se ran e sar d each othey rb a hydrostatic generator outside the chamber. Figure 5 shows a schematic diagram of the chamber and its affiliated system. studyr Inou contaca , t pressure, ambient temperatur holdind ean g time, which significantly affect the diffusion welding, are selected as a parameter. Besides these parameters ambiene ,th t pressur determinees i possibls a w lo tha o impuritee s s d a th t y composites are not included between two plates. The ambient pressure is set to 6x10"3 Pa by vacuum pumps in this test. After the diffusion process, those plates are removed from the chamber and their surface observee sar optica n a y db l electron microscop scannina d ean g electron microscope(SEM). Figur showe6 enlargee sth d cross section betwee opticae plateo th n tw y sb l electron microscope. Voids between two plates decrease as the contact pressure increases fount voidno e de Th .swhe ar contace nth t pressur mores i e than 49MPa. Effect f ambienso t temperatur weldine th significanto ee t gar ambiene th s A . t temperature changes from 1100°C to 1150°C, voids and inclusions between two plates disappear. However, when the ambient temperature is more than 1150°C, the

~~Table 2 Major specification of the new CHX~~

Type Plate heat exchangers with diffusion welding Size 170mmx 90mm* 49mm Material Hastelloy XR Design pressure S.OMPa Differential pressure betweed an w nlo S.OMPa high temperature fluid Number of stages 10 Inlet temperature Maximum 950°C

~~171 100 .(48.7)~~

~~90 6.35 36 6.35~~

~~o lO lh Mechanical V///A V7777\ . booster pump~~

~~Uppe| m ra r Cryo pump Test -^ Hydrostatic piece generator 7777\ Oil rotary pump~~

~~R0.1 -R0.5 P7771 Carbon heater Chamber~~

~~in zmrm~~

~~5.5~~

~~Figur e4 CrossX sectionaCH w ne l e vieth f wo Figure 5 Schematic diagram of the chamber and its affiliated system X400~~

~~X400~~

~~Figure 6 Cross section of two thin plates by the optical electron microscope~~

deformatio holdine n th occur platesr edge e fo th g th f s t etimeso holdinA a e . th f i , g time is ove minutes0 3 r deformatioe th , t negligibleedge no th t s e i na deformatioe Th . n deteriorates not only the thermal performance but also structural integrity for long duratio f operationsno . This results prove thaambiene th t t temperatur f 1150°Ceo , contact pressure of 49MPa and holding time of 30 minutes are selected as the best conditio diffusioe th r nfo n welding. In the next tests, test specimens of the diffusion welding will be manufactured and tensil creed ean p strength wil obtainede b l Electroe Th . n Probe Micro Analyzer analysis wil alse b l o conducte investigatdo t e inclusion diffusioe th n si n welding.

173 4.4 Thermal performance of the CHX installeloor s i ai obtai po X Tht n a ethermaCH s dn i it l conductance. Figure7 show flos a tes r wai tshee e loopth f compressoA o t. r transfer temperaturw slo r inteai o CHXe th . Afte rmaximu e that heates th i t o i ,t p dmu 120° electrin a Cn i c heate flowd ran s back to the CHX. The heat is transferred from high to low temperature air in the CHX. An s exhaustereleasei e atmosphereX th CH o t dr fro ai e d mth . Four K-type thermocouples are installed at each inlet and outlet nozzle of the CHX, and the flow rate of air is measured by a vortex flowmeter. As a result, the thermal conductance is calculated by an inlet and outlet temperatures, flow rate and specific heat of air. Tabl show3 e e resultthermae th sth f o s l performance thermae testTh . l conductance in the CHX obtained by this experiment is approximately 10% higher than analyticae thath f to l result pressure Th . e experimen e droth f po 7.3s i t % higher than that analysise ofth mors i . X eInterna turbulenCH e flor th ai lw n i t than expected because eth shoro s s i t tha lengt floX e flor ai th wtw CH f h patho cannoe th n si stablee b t . Optimizatio heaf no t transfer characteristic beet no ns carriesha becaust dou e eth focal point of this study is to develop the diffusion welding. The thermal conductance of doeX thit mee requiremense no CH th t t fro GT-MHTGRe mth . However thermae th , l conductance is sufficiently improved by modification of the plate shape or installation of turbulent promoters in the flow paths. For example, a modified CHX where a twisted tape[5] is installed in the flow paths is assumed and its thermal conductance is evaluated. Figure 8 shows relationship between the effectiveness and the CHX height. The modified CHX with diffusion welding can achieve effectiveness of 95% and its size meet the requirement (1.5m/width x l.5m/length * 5.5m/height/unit * 6 units). The CHX with diffusion weldinpossibilite recuperatoth e - s th usee GT b gha r de o fo yt th n i r MHTGR. The CHX with the diffusion welding would also become feasible for the IHX because the strength of the diffusion welding in high temperature condition is apparently higher than that of brazing, and the thermal performance and cost are superior to the HCIHX. Electric heater

~~Vortex flowmeter~~

~~Atmosphere~~

~~Compressor Difference pressure transmittem r transmitter~~

~~Atmosphere (¥) Thermocouple~~

~~Figure 7 Flow sheet of an air test loop~~

~~174 Tabl e3 Comparison between experimenta analyticad an l l thermal conductance X pressurd CH an w ne e e droth f po~~

~~Amount of Heat transfer: 533W Inleoutled an t t temp f loweo , r fluid: 23.9°C, 66.0°C Inleoutled an t t temp f higheo , r fluid: 116.0°C,73.7°C~~

~~Items Experiment Analysis Error Thermal 66.8 W/m2K 74.2 W/m2K 9.8% conductance Pressure drop 5.9 kPa 5.5 kPa 7.3%~~

~~I I I I Modified CHX~~

~~0.95-~~

~~05 §0.9 CD PresenX CH t 0.850 - 111~~

~~0.8~~

~~_L 0 1 8 6 4 Height of the CHX for the recuperator (m) maximuSize th e: m width 1.5mx1.5m Heat Capacity: 78.4MWx6units Reactor Power: 450MW~~

~~Figur e8 Relationship betwee effectivenese nth X CH heighd e sth an f o t~~

175 5. Concluding Remarks It is confirmed that the CHX with diffusion welding is one of the best candidates for the IHX in the future generation HTGR and the recuperator in the GT-MHTGR. earle Th y tests prove thaambiene th t t temperatur f 1150°Ceo , contact pressure of 49MP holdind aan g minute 0 tim3 f eo selectee bessar e th t s dconditioa e th r nfo diffusion welding weldine Th . g surfaces between plate soune sar thidn i s casee .Th tensile strengt creed han p strengt weldine th f ho g wil testee b l d this year. As for the thermal performance, the thermal conductance would meet the requirement fro GT-MHTGe mth modificatioy Rb e platth f eno shap installatior eo f no turbulent promoter floe wth n spathsi .

~~REFERENCES~~

Tanak. T al.,"Constructiot a] e [1 Testin s HTTe it th d f nRo gan Program", presented workshod an M "Desigan ptTC o Developmend nan f Gas-Cooleo t d Reactors with Closed-Cycl Turbines"s eGa , INET, Tsinghua University, Beijing, China, 30 Oct Nov.2 .- , (1995).

Kunitom. K [2] t al.,"Strese i Straid san n Evaluatio f Heano t Transfer Tuben si Intermediate Heat Exchange HTTR"r rfo , ASME/JSME Nuclear Engineering Conference, 847-853, San Francisco, March (1993).

~~[3] S. Sato, Private communication, (1995).~~

~~] [4 GCRA, "EvaluatioTurbins Ga e eth Modulaf no r Helium Reactor", DOE-HTGR- 90380, Dec. (1993).~~

~~. LidskyM . L . Mor ,d M privatan i[5 ] e communication, (1990).~~

176 SUMMARY REPOR TECHNICAN TO L EXPERIENCES FROM HIGH-TEMPERATURE HELIUM TURBOMACHEVERY TESTING IN GERMANY WEISBRODIA T* ,,..,,,.,..,,,™™__. Nuclear Power Technology Section, Division of Nuclear Power, IAEA, Vienna

~~Abstract~~

In Germany a comprehensive research and development program was initiated in 1968 for a Brayton (closed) cycle power conversion system. The program was for ultimate use with a high temperature, helium cooled nuclear reactor heat sourc projectT e electricitHH r (thfo )e y generation using helium as the working fluid. The program continued until 1982 in international cooperation wit e Unitehth d State Switzerlandd san . This document describe designe sth reportd san e th s result f testinso g activities that addresse developmene dth turbinesf to , compressors ductss ga t ,ho , materials, heat exchangers and other equipment items for use with a helium working fluid at high temperatures.

The program involved two experimental facilities. The first was an experimental cogeneration power plant (district heating and electricity generation) constructed and operated by the municiple utility, Energieversorgung Oberhausen (EVO) t Oberhausena , , Germany t consisteI . fossia f do l fired heater, helium turbines, compressor related san d equipment. secone Th de facilitth s ywa High Temperature Helium Test Plant (HHV) for developing helium turbomachinery and component Researce th t sa h Center Jiilich (KFA) derive V heae Th HH t. sourcde th fro r n mefo a electric motor-driven helium compressor. These facilities are shown in the photos on the following pages.

In both facilities negative and positive experiences were gamed. At initial commissioning operation difficulties were encountered with the EVO facility, including failure to meet fully the design power output of 50 MW. The reasons for these difficulties were identified and as far as economically feasible the difficulties were corrected. At the HHV some start-up problems also occurred wert bu , e soon corrected researce Th .developmen d han t program t botsa h facilitien ca s be judged successfu fulld an ly supportiv higf o he feasibilite temperaturus th f e eo th f yo e helium as a Brayton cycle working fluid for direct power conversion from a helium cooled nuclear reactor.

Unfortunately projecT terminates HH wa te th , Germann di botd yan h test facilities have been shut down. Except for information on life testing the facilities accomplished their missions. If helium turbomachinery technolog reconsideres yi power dfo r conversion fro mheliua m cooled nuclear reactor unresolvablo n , e problems have been identifie thesn di e turbomachinery test facilities.

* Present address Loner Hohenweg 22, D-52491 Bergish Gladbad, 1 i Germany

~~177 1. Introduction~~

This report give e a desigsurves th d operationa f an o ny l experiences gained from high temperature, helium turbomachlnery testing in Germany, which formed a portion of the high temperature turbine (HHT) development program. In addition appendices have been provided on other experimental program shora T projecn T o tsHH histord relevane HH an tth e f th y o o t t project (Appendi. xI)

~~This report describe initiae sth l difficulties improvemente th , resultse madth d sean achieved.~~

Some of the experimental results achieved are not publicly available, because they are classified as "proprietary information" within the gas-turbine-technology and high-temperature-technology industrial communities of Germany. Therefore in some cases only a short survey could be given.

~~2 Desig f Tesno t Facilitie Descriptiod san f Turbomachinerno y~~

2 1 Helium Turbine Cogeneration Power Plant (EVO)*

~~21.1 Development Goals~~

~~(Ref. [1], [2], [3])~~

Beginning in 1960 the municipal energy utility of Oberhausen (EVO) operated on its grid a closed-cycle, hot-air turbin ecombinee planth r fo t d generatio f electricitno districd yan t heat. increasino t e Du g demand extension a , powee th f no r plant capacit decides ywa d upo 1971n i .

~~EVO: Energieversorgung OberhauseG nA~~

~~OBERHAUSEN II HELIUM TURBINE PLANT~~

~~178 HIGH PRESSURE ROTOR FROM OBERHAUSEN II HELIUM GAS TURBINE~~

~~ROTOR FROV MHH TEST FACILITY~~

179 The EVO decided to apply a closed-cycle helium turbine cogeneration power plant in consideratio requiremente th f no commercialla r sfo y operated powe cogeneratiore planth r tfo n of electricit districd yan t heat. This decision recognize favorable dth e thermodynamic properties of helium, which promised an attractive application of closed-cycle gas turbines, as well as the goals for the international development of the High Temperature Reactor (HTR) using a direct cycle High Temperature Helium Turbine (HHT) power conversion systemh 4t defines a ,e th n di Atomic Program of the Federal Republic of Germany. With this plan a large-scale practical contributio Germae th o nt n nuclear progra commercialla mn i y operated power plant coule db made. In order to optimize the experiences to be gained, it was decided to provide a "nuclear design" for the helium-systems and the turbomachinery and to use as high helium-temperatures and helium-pressures to the greatest extent possible, taking into account the staterof the art of material componentd san decisioe th t sa n date.

~~2.1.2 Test Operation Goals~~

~~(Ref. [3]) followine Th g experiences gainewere b o et d wit tese hth t plant:~~

~~• Nuclear design of the helium systems and of the turbomachinery~~

• Mechanical, thermal, aero-dynami dynamid can c behaviou helium-turbine th f o r d ean low-pressure oth f e (LP high-pressurd )an e (HP) compressors, including rotor shafts, blades, shaft seals and cooling systems

~~• Acoustic emissions and propagation within the plant and into the environment~~

~~• Behaviour of hot gas ducts, bellows, valves and insulation under thermal loads, pressure gradients, acoustic emission heliud san m impurities~~

~~• Experience with different operating conditions (start-up, steady state powe shutd an r - down)~~

~~• Power regulation~~

~~• Remote operation (simulating nuclear reactor applications)~~

~~• Behaviour of all other helium-components including heat exchangers and purification system (except radioactive impurities)~~

~~• Longterm operational behaviou heliul al f ro m component heliud san m system wels sa l turbomachinere th af so y unde conditione rth commercialla f so y operated, cogeneration plant~~

~~180 2.1.3 Desig Systed nan m Description~~

~~(Ref. [ 4 ], [ 5 ], [ 6 ], [ 7 ], [ 8 ], [ 9 ], [ 10 ], [ 11 ], [ 12 ], [ 13 ], [ 14 ], [ 15],[ 16 ])~~

~~) a Overall Design~~

The design of the EVO test plant provided for an electrical power of 50 MW and a heating power (district heat) of 53.5 MW. The inlet temperature into the turbine was limited to 750 °C, taking into accoun life th et enduranc materiale th firse f th eo t r turbinsfo e stage f bladeo d san the helium heater.

A two-shaft desig selectes turbomachinerye nwa th r dfo highe Th .- pressure turbine drivee sth compressors on the first shaft and the low-pressure turbine drives the generator by a separate shaft. Both rotors are interconnected by a gear.

applies Aiworkine wa rth s da g l previouslmedial r afo y operated closed-cycl s turbineega t a s EVO, because the physical properties of air had been well known. Compared to air, helium has distinct advantages for a closed-cycle gas turbine. First of all it is a chemically inert gas. Moreover, it has very advantageous physical properties as can be seen from Table 2.1./I.

~~Table 2.1./1: Physical Properties of Air and Helium~~

~~Air Helium Sound velocity at C ° 0 2 s m/ 3 34 100s 7m/ C ° 0 60 584 m/s 173s 8m/~~

~~Specific t heaa p c t C ° 0 2 r ba 1 1010 J/kg K 5274 J/kg K C ° 0 60 r ba 0 3 1120J/kgK 5274 J/kgK~~

~~Isentropic coefficienC ° 0 2 t r ba 1 1.40 1.665 30 bar 600 °C 1.36 1.665~~

~~Heat conductivity 1 bar 20 °C 0.026K m 5W/ 0.146K m 6W/ 30 bar 600 °C 0.0628 W/m K 0.3287 W/m K~~

~~Dynamic viscositC ° y0 2 r ba 1 1.82 lO'5 kg/ms 1.98 10-5 kg/ms 30 bar 600 °C 3.87 10-5 kg/ms 4.18 lO'5 kg/ms~~

e specifiTh c hea f heliuo t measurma ( heae th tr capacitefo head yan t transport s fivi ) e times larger than tha f airo t , thus requiring smaller heat transfer areas soune Th .d velocit thres yi e times as large, resulting in design advantages for the turbomachinery. In particular, the permissible circumferential velocit heliumcompressoe th f yo longeo n s ri r limite soniy db c speed consideration but by the centrifugal stresses of the blades. A characteristics of helium's large specific heat is that the enthalpy difference between preselected temperatures is accordingly large, resultin largea n gi r numbe f stage heliuo ra r r turbinesfo ai m n a tha r .nfo

~~181 b) Flow Scheme~~

~~The basic flow scheme of the EVO helium turbine plant is shown in Fig. 1.~~

The helium working media is initially compressed in the low pressure compressor (1), recooled in the intercooler (2) and additionally compressed to the plant design pressure of 30 bar in the high pressure compressor (3). Preheating occurrecuperatee th finan d si an l ) heatin(4 re th o gt plant design temperature of 750 °C occurs in the gas-fired heater (5). The compressed and heated helium then expands in the high-pressure (HP) and low-pressure (LP) turbines, (6) and (9) respectively compressioe Th . n ratio betwee low-pressure nth high-pressurd ean e compressors was selected in such a way as to optimize the heating power for district heating, which is transferred within section (8.1precoolere th f )o , while section (8.2) further cool heliue sth m working fluid prior to its entering into the low-pressure stage of the compressor (3).

~~8.2 8.1~~

Legend: pressurw Lo e compresso1 r 7 Low pressure turbine 2 Intercooler 8 Precooler 3 High pressure compressor 8.1 District heat removal section 4 Recouperator 8.2 Precooler section 5 Heater 9 Gear 6 High pressure turbine 10 Regulation bypass

~~Fig. 1: Flow-scheme~~

~~182 thermodynamie Th c cycle point condition plane showe th r ar t sfo Tabln ni e 2.1/2.~~

~~Table.2.1/2: Thermodynamic Cycle Point Conditions Data~~

No. Component TemperaturC e° Pressure bar 1 Low Pressure Compressor 25 10,5 2 Intermediate Cooler 83 15,5 3 High Pressure Compressor 25 15,4 4 Precooler High Pressure Side 125 28,7 5 Heater 417 28,2 6 High Pressure Turbine 750 27,0 7 Low Pressure Turbine 580 16,5 8.1 Heater Inlet 460 10,8 8.2 Cooler Inlet 169 10,6

e inleturbinP Th H tr s selecteint e pressurba oewa th compromisea 7 2 s d a f o e e largesTh . t pressur closer efo dthat a cyclr t turbines timeba ega 0 4 . s sCondition witwa r hai nucleaa r sfo r heated plant provided with a helium-turbine were foreseen as about 60 bar, a helium-temperature of 850 °C and a plant capacity of 300 MWe. The low pressure turbine was selected to deliver power at a frequency of 50 Hz or 3000 rpm and is connected by a gear arrangement to the high pressure turbine shaft that rotates at 5500 rpm. It has dimensions and stress loadings for the e turbinrototh r e bladefo er th shafhousind r an sfo t g whic e verhar y simila thoso t a r f eo 300 MW helium-turbine plant considered at the tune as the reference size. In addition to optimization requirements a reason for the limitation of the primary pressure was the permissible stressemateriae fossil-firee tubeth e th r th f fo sf so o l d heater.

~~2.1.4 Turbomachinery Description~~

~~) ] 5 1 [ , ] 4 1 [ , ] 3 1 [ , ] 2 1 [ , ] 7 [ , ] 6 [ , ] 5 [ , ] (Ref4 [ .~~

e roto compressoP Th L r e shaftth f o s r commoe (seth d e Figan n ) 2 roto.P H re shafth f o t compresso turbinP H d ean r (see Fig ) wer3 . e couple geaa y drb assembl e nominath r fo y l revolutioshafP H 550 s i te compressoP th 0 L f nrpmo e Th locate.s separats rwa it n di e housing and the HP compressor and the HP turbine were located within a common housing. The split-up of these three machines into two housings was made to avoid natural frequencies of the rotor shafts at run-up, operational speed and rundown.

In orde ensuro t r easn ea y accessibilit bearinge th o yt s betwee compressoP P L H e e nth th d an r compressor without requiring the removal of the main housings the two housings were interconnecte pressure-tigha y db t tunnel.

As shown in Fig. 2, a ball-shaped housing, horizontally divided and welded from 8 ball-shaped segments, was provided for the LP compressor. Thus led to minimizing both the housing wall thicknes e helium-leath d an s compressoP L k stages0 e ratee 1 bladin Th d Th s . .ha r gwa designed according to the potential vortex theory. Along the complete blade length a reaction ratio of 100 % was provided.

~~183 (0 ts o O d. -J~~

184 oo Fig. 3: HP-Compressor and HP-Turbine As show commoa , Fign ni 3 . n housin P compressoP H H provides e e gwa th th d r dfo an r turbine, simila provisione th provee r turbino t rai th t r nfo sho e plants naturae vien th I .f w o l frequencies rotocompressoe P th H , re shafth provides f o t wa r d wit boreha , wherea rotoe sth r turbinP H forme s e oeth fwa singly db e disks connecte Hirth-typa a dvi e coupling. Both rotor sections were firmly bolte eaco dt h othetube-shapea y rb turbinP dH boltdesige e eth Th .f no housing was made with regard to the design of large-scale turbines for nuclear heated plants. Therefor inneo n e r insulation t foili r fibrese o sb s provided , wa , . Rathe a separatr e inner turbine housing was provided, which was cooled by the passage of outlet helium from the HP compressor.

~~The common housing of this unit is made of 8 ball-shaped segments at the ends and a cylindrical center section.~~

The HP compressor has 15 stages, which like in the case of the LP compressor are designed for turbinP stag7 H s e ae ha reactio Th bladin . % n0 g ratiwit10 f ohreactioo a r fo % n 0 rati5 f oo the medium blade-length of the final stage. The turbine rotor is provided with cooling for the rotor disks and for the blade feet. Since at nominal power the HP turbine power is just sufficient to drive the LP compressor as well as the HP compressor, the gear interconnected between HP section and LP turbine transfers only a small amount of power.

Fig show4 . 11-stage sth turbinP eL e provided wit hballshapea d housing also mad ball8 f e-o shaped segments. In accordance with the prevailing grid frequency of 50 Hz the LP turbine operate 300t housinde a outlee th th 0 f d inlee o rpmt gan p Th tlocateto . s nozzle wa th ) t da e(a nozzle (b) was at the bottom, thus minimizing the forces on the connecting pipes. Like in case turbineP H e oinneo th fn , r insulatio provideds nwa . Howeve outen ra r insulatio outee th rn no housing made of mineral wool was provided. The inner housing as well as the inlet nozzle and inlet housings, located insid outee eth r housing, were cooleoutflowine th y db g heliume Th . inlet sectio housine stationare welth s th a f ns o ) a l g(d y blade carrie wer) r(e e accurately adjusted by mean f severaso l screw bolt locate) outee s(f th n rd o housing orden I . ensuro t r betteea r thermal insulation the stationary blade carrier (e) was enclosed by a thin sheet-plate envelope (g).

For the sealing between bearings and the helium-circuit, a seal gas system consisting of 3 labyrint chosens wa ) hk . seal, Intj chambee , os(i th , provide2 r d betwee labyrite nth h seal) s(i and (j) sealing helium was introduced. (In the case of a nuclear plant, clean helium would there be introduced) e labyrintth n I .helium-floe h th sea) (i l s directewwa d towar e bearinth d g (chamber 1). The sealing helium was mixed there with the oil drained from the bearing and was subsequently regaine oil-helium-seperation a i dh n system.

Helium fro primare mth y directesids ewa d into chambe mixed an 3 dr there wit sealine hth g helium flowing from chambe througr2 labyrinte hth h seal (i). From ther flowet ei d througe hth labyrinth sea) int primare (k lo th y helium-system.

By mean adequatn a f so e regulatio pressure th f no e levels withi chambere nth s provided between the labyrinths seals it was ensured (for future application to a nuclear heated system) that under no condition operatiof so n would primary helium carry radioactive impuritie circuitl s oi int e oth . In order to seal the shaft penetration at the housing, a floating ring seal and a static seal were provided. This sealing principle was used for all 3 turbomachines.

The common roto compressoP turbinrP H H shaf d f o an ets supporte bearingo wa r tw n o ds provided with multiple plain surfaces compressoe Th . r sectio rotoe th f rn o shaf mads wa f t eo one forged piece using low-alloyed steel. This semi-finished forging was then machined to a

186 Legend: a) Inlet nozzle b) Outlet nozzle c) Outer housing d) Inlet housing e) Stationary blade carrier /] Screw bolts g) Sheet plate envelope h) Radial bearing ) k , I) , /) 3-stage labyrinth seal /) Floating ring seastillstand an l d seal 1, 2, 3 Labyrinth chambers

oo Fig. 4: LP-Turblne drum-shaped rotore turbinTh . e rotor shafs mad wa f high-alloyet eo d austenitic steel disks connected by means of a Hirth-type coupling and screw bolts to form the turbine rotor. HP compresso rotoP r H roto rd weran r e both connecte screy db w bolt foro t se complet mth e common rotor shaft. In spite of the provided blade foot cooling, austenitic steel was chosen for the turbine rotor in order to ensure an appropriate behaviour in case of all credible operationing conditions.

~~The blades for the HP compressor were made of forged semi-finished parts, which subsequently were precisely machined HP-turbinee bladee th Th r . sfo s were mad precisioy eb n forging.~~

Due to the high sonic velocity of helium, it was possible to provide higher circumferential speeds without closely approachin sonie gth c range. Therefore possibls wa t i ,desigo et e nth compresso same th t ea r circumferentia reactio% 0 l10 n spee a ratio r dfo , resultin highea n gi r enthalpy increase step tha cascompressoa n ni f eo r stag reactio% e wit0 5 hna ratio.

~~The bearing housings of all three turbomachines could be removed separately without opening the horizontal flanges of the main housings. The flanges were provided with welded lip seals.~~

cooline housinge th th r three f gth o Fo ef so turbomachines multipla , e housing system witha flo f low-temperaturwo e heliue outeth t mra sectio s providenwa orden i d o avoit r d inner insulation ordeturbinn P I .H coo o t re eth l roto bladd an r e feet coolina , g flo f heliuwo s mwa extracted from the HP compressor outlet and guided through the bore of the rotor shaft to the ring-shaped chambers of the turbine rotor disks and then finally, after the last turbine stage, ducted intmaie oth n helium-stream e ring-shapeTh . d chambers were formee outeth y rb d surfaces of the rotor disks and the blade shoulder.

~~2.1.5 Power Regulation~~

~~) ] 5 [ , ] (Ref4 [ .~~

The power regulation of the closed-cycle, helium turbine plant was designed using the same principle closed-cycle s th use r fo d r turbinai e e e normaplantsth n I l. case poweth es wa r regulate meany b dpressura f o s e level regulation. When lowerin e loadgth , helius mwa extracted from the main stream after the HP compressor through a regulation valve into storage tanks. In case of a load increase, helium was returned from the storage tanks through an inlet valve arranged ahead of the LP compressor into the main system.

planr Fo t operation prio grid-synchronizatioo rt regulatioe nth machine th f no e revolutions occurs by means of a regulated bypass line provided between outlet of the HP compressor and outlet of the LP compressor. Thus the high-pressure section of the heat exchanger and the helium- heate e turbine s th wela rs a l s woul e bypassedb d . Sinc e requireth e d compressor power remained nearly constant, only abou circulatee thira tth f do d helium mus bypassee b t orden di r to achieve zero net power output.

~~188 2.2 High Temperature Helium Test Plant (HHV)~~

~~2.2.1 HHV Goals~~

(Ref. [17], [18], [19]) integran a e b developmene lo t th planpar V s f o twa HH t e Th t projec high-tempertura r fo t e reactor wit direct-cycleha , helium turbin largf eo e capacity (HHT). This projec carries wa t d out in an international cooperation between the Federal Republic of Germany, Switzerland and the United States.

Although proven gas turbine technology was used to the largest possible extent, the development s showha T n thaworHH r t experimentakfo componentT vitad an HH l w ne le stestth werf so e required. Since no test facility of sufficiently large size was avaible, it was decided to set up a new test facility at the Research Center Jiilich (KFA). This should meet the HHT requirements with regard to the necessary test conditions (sufficiently large helium flow, high helium temperature). 2.2.2 Test Operation Goals

~~(Ref. [ 16 ], [ 17 ], [ 18 ])~~

~~componentT teso t planHH tV s e goa wa Th HH tl f sufficientlo s y large siz permio t e e th t extrapolation for HHT use. followine Th g test loop conditions were specified:~~

~~• Helium mass flow:approximately 200 kg/s~~

~~• Helium temperature: 850 °C with the possibility to reach 1000 °C for short time periods~~

~~• Helium operating pressure:50 bar~~

~~• Adequate helium atmosphere with regard to non-radioactive impurities~~

~~• Sufficiently large test section for employing sufficiently large components to enable an extrapolatio HHTe th o .nt~~

Tes followine th t r resultfo gV s maicomponent T e achievewerHH b ne HH o et th n di d an s systems: • Helium turbine with gas seals, oil seals, bearings, cooling systems, insulation • Helium compressor with gas seals, oil seals, bearings • Helium/helium heat exchangers/recuperators, coolers ducts ga t s witHo h valves • , butte valvesy fl r , bellows, bends • Helium purification • Coating heliun si m atmosphere • Material behaviour in helium atmosphere at high temperatures • Instrumentation • Acoustic emission propagationd an s s • Dust and particles in the helium-circuit

~~189 2.2.3 Design and System Description~~

~~) ] 8 1 [ , ] 7 (Ref1 [ .~~

a) Basic Design e firsTh t design considerationplanV tHH begae th r n fo swit hsystema , wher fossil-fireea d heater replaced the nuclear heat source. However, because of the desired peak temperatures of 850 °C (1000 °C for shorter periods) and the design pressure of 50 bar mere were feasibility concerns abou lifetime th t e capabilit f sucyo hfossil-firea d heater. Thu tesa s t circuis wa t chosen comparabl closed-cycla o et turbins ega e plan shows a t Fign ni . 5 .

In this arrangemen compressioe th t n heat fro compressoe mth uses heliue i r heath o d t p mo u tt the desired temperature. The required compressor power was 90 MW with 45 MW generated and regained by the expansion in the gas turbine and 45 MW introduced by the electric drive motor. The selected combination of the turbine and compressor on one shaft resulted in dimensions being comparable with a helium turbine of 300 MW capacity (the reference plant size at the time).

~~b) Design Description~~

~~The flow scheme of the HHV test loop is shown in Fig. 6.~~

~~Heliusystes ga circulate s t mean electrically-drivey mme b ho wa th e f so th n di n turbomachinery. A synchronous-moto electricae uses th s dwa a t 300 a r m l 0rp drive .~~

~~1 Test Section~~

~~Legend: 1 Test section 3 Duct to compressor inlet 2 Helium-turbine 4 Compressor~~

~~Fig. 5: HW test circuit~~

190 Legend: 1 Turbomachinery 2 Synchronized W motoM 5 r4 3 Start-up motor 4.5 MW 4 Cooling gas compressors 5 Motor for cooling gas compressors, 6.5 MW 6 Hot-gas duct 7 Bypass Coaxia8 l test section 15 9 Test section vessel 10 Hot-gas outlet 11 Emergency relief 12 Cooling gas inlet 13 Heater 14 Main cooler 15 Intermediate cooling water circuit 16 Cooling water pump 17 Water-air cooler 18 Sealin cooles gga r 19 Hot-gas cooler 20 Helium storage 21 Piston-type compressor 22 Vaccum pump 23 Helium purification system

Fig. 6: Overall flow scheme of the HHV The highest helium-temperatures were compressoP achieveH e th t da r outlet t heliuHo . m could be conducted completely or partially through the test section or directly bypassed back to the turbine for expansion by means of hot gas ducts provided with regulation valves.

~~A helium/water cooler was provided in order to ensure the desired temperature equilibrium between adde removed dan d hea supplo wels t a t s a l y sealincoolind an s ggga gas.~~

~~The principal design parameters for the helium circuit are given in the following Table 2.2/1.~~

~~Table 2.2/1: Principal Design Parameters of Helium Circuit at Nominal Power~~

~~Temperature Pressurr eba Flow Rate °C kg/s~~

Compressor outlet 850 51.20 212.0 Test section inlet 850 51.20 201.0 Turbine inlet 826 49.70 209.0 Main cooler inlet 390 49.50 53.5 Cooling gas compressor inlet 236 49.00 56.8 Coolin ducs r coaxiafo ga tt s gho ga l 300 51.20 22.9 Sealin outles gga t 50 52.75 2.3

~~The total helium conten circuie th approximatels n i twa t heliue th d my an 800throughpu3 0Nm t for purification was approximately 1000 Nm3/h.~~

~~2.2.4 Descriptio Turbomachinerf no y~~

~~(Ref. [ 17 ], [ 18 ])~~

~~As show compressoe turbine th th Fign nd , i 7 ean . r were arrange singla n do e shafte Th . turbine has two stages, the compressor eight.~~

Helium flowed through the inlet nozzle into the turbine section. Behind the compressor outlet helium flowed throug outlee hth t diffusor intoutlee oth t nozzl frod ean m s therga t e ho int e oth duct. Insid turbomachinere eth y onl inlee outled yth an t t nozzles diffusore bladine th ,th d gan s channel were exposed to high temperatures. All other sections of the machine (blade feet, rotor and housing) were coole meancooliny e db th f systes o sealin e gga th msyster s o g ga m (seg eFi 8).

The rotor provided for the turbomachine had a weight of 60 Mp, a bearing distance of 5.7 m and a total length of about 8 m. The blading channel had an inner diameter of 1.6 m and an outer diameter of 1.8 m at an overall length of about 2.3 m.

The rotor was constructed by welding several forgings of heat-resistant ferritic material and was provided with longitudinal groove r attachinsfo bladinge gth turbine Th equippes . ewa d with two blade rows, each consisting of 84 rotor blades and 90 stationary blades, while the

~~192 5324 1 7. 2 7~~

Legend: 1 Rotor 6 Stationary blade carreir 2 Outer housing 7 Blading 3 Bearing 7.1 Turbine section 4 Labyrinth seals 7.2 Compressor section 5 Shaft seal 8 Connecting nozzle frod man shot-gao t s ducts

~~: FigLongitudina7 . l sectio turbomachinf no e Legend:~~

1 Inlet at stator compressor section 4 Low pressure side inles t Ga nozzl 7 e 10 Sealin outles gga t 13 Sealing gas inlet 2 Sealing gas outlet 5 Sealing gas inlet 8 Outle statot ta r 11 Outlet from rotor outles 1Ga 4t nozzle 3 Inlet into rotor 6 Gas outlet nozzle Inle9 turbint ta e section 12 Drive-motor side inles 1Ga t5 nozzl e

Fig. 8: Cooling and sealing gas flow in the turbomachine compressor had 8 blade rows each with 56 rotor blades and 72 stationary blades. All blade feet were provided wit hheliua m cooling bladee Th . s were manufactured from NimocasC L 3 71 t using a vacuum precision casting procedure.

The rotor was supported on two segmental shoe bearings with forced oil lubrication. The one bearing located outside the turbomachine between drive motor and turbomachine at the interim shaft was designed as the axial fixed point of the rotor shaft (thrust bearing). The fixed point turbomachine foth r e housin provides gcompressowa e th t da r outlet stud. turbomachine Th e housin mads gwa cas f horizontall es o t wa stee d an l y divide flanga t da e joint. The flange surfaces had been growned and a sealing paste (Poly-Butyl-Cuprysil) was used to ensure the tightness. However, at the first leak tests it became evident that this sealing was not sufficient. Thus three additional grooves for accommodation 0.3 mm diameter gold rings and a purge groove were provided (see. . Fig9) . e schematiTh c design shafe th f t so seal r operatioe bearin fo sr shutdown th fo r d fo g nan d an , lubrication oil supply are shown in Fig. 10. The shaft seals are connected to the front flanges of the turbine housing. chosee Th n floating ring seal design correspond provee th o st n technology use hydrogenr dfo - cooled generators. In addition, this design has been pretested at Brown Boverie Cie (BBC) - now Asea Brown Boverie speciaa (ABB n i - )l seal testing stand.

~~2.2.5 Power and Temperature Regulation~~

~~(Ref. [ 17 ], [ 18 ])~~

The power and temperature regulation of the test plant occured at constant revolutions of the turbomachinery temperature Th . pressure wels th e a s la e were regulate adjustiny db heliue gth m flowrate, respectivel cooline yth g water massflo coolee heliue th th t wn a i mr bypass line (see Fig. 5).

~~The nominal operating temperature of the helium main circuit could be regulated between 850 °C and 1000 °C. The nominal pressure of 50 bar remained constant.~~

~~3. Technical Experiences gained fro Tese mth t Plants~~

General Remarks e technicaTh l experience tesV st gaineHH facilitie d an de summarize ar O fros mEV d from numerou expectee b so t reportss d wa tha judgementt e I .th t differenf so t companie personr so s same onth e subject woul alwayt dno consistente sb largese Th . t portio thosf no e discrepancies might be explained by a different background of experiences or different technologies known and applied by the different companies and persons. Nevertheless it was attempted to select the most competent judgements on the specific items. But hi case of the most important items (e. g. the turbomachinery) different judgement r statementso r detailo s retainee sar d with respece th o t t same subject in order to ensure that no important aspects were omitted.

~~195 Legend:~~

~~7 System pressure 4 Outer atmosphere 2 Purge groove 5 Bolt chamber 3 Sealing wire~~

~~flangV HH e : joinFig9 . t seal~~

~~196 Legend:~~

1 Helium sealing oil 4 Air oil storage tank 7 Helium separation tank 10 Helium/bearinl goi Vacuu2 m sealinl goi 5 Vacuu l storagmoi e tank 8 Helium sealin l storaggoi e tank 11 Helium/bearin l storaggoi e tank 3 Air sealing oil 6 Air 9 Standstill seal 12 Helium Fig. 10: Schematic designs of the shaft seals for operation and for stillstand and of the lubrication oil supply for the bearings 3.1 Turbomachinery

~~3.1.1 EVO - Turbomachinery~~

~~3.1.1.A Summary on the operational experiences as published by EVO~~

~~) ] 8 2 [ , ] 7 2 [ , ] 6 2 [ , ] 5 2 [ , ] 4 2 [ , ] 3 2 [ , ] 2 2 [ , ] 1 2 [ , ] 0 2 [ , ] 9 1 (Ref [ .~~

~~General Operation Experiences~~

turbogeneratoe Th initialls wa r y synchronize grie Novembed n th do en o dt e th , o 1975t 8 r p U . of 1988 (final shut-down) the helium-turbine plant had achieved approximately 24,000 hours of operation, not including periods when it was used for test purposes in isolated operation. A total of 11,500 hour f operatioso desig e beed th nt ha n a n temperatur. °C 0 75 f eo During the operation many components and systems showed good performance. However, for some components unexpected problems arose.

~~Vibration Problems~~

Since it was known and expected that the long and slender rotor of the HP-set would be susceptible to vibrations, the vibration behaviour was carefully precalculated in the planning phase dynamie Th . c behaviou roto e distinguishes th f wa r o r followss da :

~~• forced oscillations (du unbalanceeo t s and/or thermal distortions)~~

~~• resonance frequencied an s~~

~~• self excitated oscillations (sealin excitationsp ga g , hydromechanical e forceth n i s lubricatio l filmnoi , elastic hysteresis, shrin t frictionkfi )~~

The first resonance frequency of the HP-set was originally calculated to be about 2050 rpm, assuming stiff bearings firse Th t . resonance frequenc LP-see th calculates f yo wa t aboue b o dt t 1800 rpm.

These calculations indicated that a large safety margin for the stability limit was to be expected. Nevertheless firse th t t pressur e a startu,soos th a s temperaturnd d a pan ean beed eha n increased a sudden rise of the shaft oscillation amplitudes at the rotor of the HP-set was detected. These oscillation amplitudes wer largo es e thadesige th t n valu f eo spee powed dan r e coulb t dno attained. Moreover severe damages occured at the sliding surfaces of the bearings including partial tearin bearine gth g surfaces.

At about 145 excessivelm 0rp y large oscillation self-excitateo t e sdu d vibrations were observed. e measureTh d first resonance frequenc s abouwa ys agains t (a e 195 th m n t0i rp 205m 0rp precalculation) conclusioe Th . , thanis mustm approximateldeductee rp b t 0 20 o dt fro0 y10 m the theoretical calculations for stiff bearings to meet the realistic conditions.

~~198 »4o— 0> T3 ^3~~

~~"5.~~

~~Time~~

~~Fig. 11: Measurements of the vibrations actually accrueing at the bearings of the HP-turbine (measuring points staggered by 90°) before performing modifications~~

In order to overcome the large oscillation amplitudes which mainly were due to large unbalances and thermal distortions of the rotor, new balancing was carried out and the gear was modified to permit a lower running speed of the HP-set. As further reasons, a gap excitation was suspected together with a "unfavorable design" of the bearings. To clarify these latter items special measurements for different types of bearings were performed.

Fig. 11 shows a record of the overall amplitude of the rotor oscillations measured at two measuring points staggere 90°y db . These measured oscillations were typicavibratione th r lfo s observed with the EVO-turbomachinery and indicated self-excitated vibrations. Taking into account the known effects that influence a self-excited vibration, various countermeasures were taken including providing transverse strips in the labyrinth seals.

Instea f orginalldo y provided cylindrical bearings various type f bearingso s , multiple plain surfaces bearings with/without grooves, different wedge-typ f bearingo e s with different width/diameter ratio varioud san s tilting-segment-type bearings, were also testecourse th i dh e of further operation.

It was also decided to modify the rotor support that had a large influence on the vibration behaviour. The rotor design was changed in order to shorten the bearing distance and to achieve a stiffer drum shaft within the compressor section. Fig. 12 shows a longitudinal section of the rotor of the HP-set before and after the modification. This change led to a substantial increase criticae oth f l rotor frequenc runnine th d gan y behaviour became very quie satisfactoryd an t . The theoretically calculated resonance frequency (assuming stiff bearings) was then approximately 2600 rpm.

The result of these efforts was that there was a distinct rise of the stability limit with regard to higher operating pressures and higher speeds (speeds increased in steps from 4532 rpm to 5136 rpm, to 5300 rpm and finally to 5500 rpm). The power of the turbo-set did not reach the nominal power, however.

~~199 5660~~

Fig.12: Modification rotoe th f r sHP-se o e shaf th f to t unexpectee Th d vibration problem severd san e bearing damages wer largea cause parth f eo t unplannee oth f d plant shutdowns. However countermeasuree th , s taken wer effectivo es e that the problem of the rotor oscillations is considered solved. After having carried out the above modification measurementw sne s wer etotae taketh lr turbosenfo especialld an tHP-sete th r yfo .

~~Fig. 13 shows the measurements of the vibrations actually accrueing at the bearings of the HP turbine (measuring points staggere 90°y db ) afte modificationse th r .~~

This figure shows distinctly thae turboseth t s runninwa t g quietl d thae oscillatioan yth t n amplitudes still observed /*morde0 e 1 th f , n rwero i , e sufficientl turbosetr fo w ylo thif so s size. Only at one measuring point was an amplitude of 110 pm measured. Overall, the vibration pattern was smooth and the machine ran quietly.

Figshow4 1 runu.e sth p spectrum measure sensoa y db r locate turbine th P t dH a ee sidth f eo shaft rotor (measuring point VO 4). This is further proof of the success of the improvements. The indicated retention time i havFig4 sh 1 .e been shortened substantially. This figureo to shows, tha oscillatioe th t n amplitude frequenciel al t sa s have become reasonabl acceptd an w -ylo able.

~~Analysis of Vibration Problems~~

The self-excitated and gap-excitated oscillation and bearing problems of the turbine revealed a surprising discrepancy betwee time nth theoryet a whe t helium-turbine ar n,th e statth f eo s ewa designed, and the actual results. It must be pointed out, however, that fundamental investigations regarding gap excitation and bearing characteristics had only been initiated at the time when the helium-turbine was under construction. Moreover, during the design the excitation forces resulting from the blading and accrueing hi the labyrinth seals could only be roughly estimated. The excitation forces accrueing in the labyrinth seals were also known only approximately.

~~CO §~~

~~> B •CoD "5.~~

~~Time~~

~~Fig.Measurements13: vibrationsthe of actually accrueing at the bearings of the HP-turbine (measuring points staggered by 90°) after the modifications~~

~~201 RMflX mn-3]~~

~~.0 .0 30.0 2500.0~~

~~5000.0 20.0 Time [min]~~

~~7500.0 10.0 10000.0 Frequency [ 1 / min ]~~

~~Measuring point4 O V :~~

~~Fig. 14: Runup spectrum of the EVO-turbomachinery (October 1980)~~

The bearing characteristics knowledg onls ewa y improved continuously fro beginnine mth f go the 1970's onward n initiate198O I . 7EV a dcomprehensiv e summarizing analysid an s judgement of the bearing problems, carried out by competent experts. The characteristics of many types of bearings and bearing geometries applied in the course of the operation of the turbine, were recalculated in terms of a permissible gap excitation coefficient, q. The

~~summarizing result givee sar . Fign ni 15 . er~~

The so-calle excitatiop dga n coefficien introduces wa t characterizo dt excitatiop ratie ga eth f oo n forcd rotoan e r bending amplitudeexcitatiop ga e Th n. coefficien e achieveb e o th t t y b d bearings, as calculated, is given on the ordinate. It can be seen that the original bearings even had a negative or extremely low qCT being the cause for the unfavorable vibration behaviour. From bearin onward5 . distinctls i g No r e sq y larger than zeromeasuree Th . s subsequently taken furthea o t d rle improvemen q e seeere th b . f nThun o t tha ca counter-measuree t si tth s taken were correct.

A theoretical calculation of the prevailing gap excitation coefficient on the basis of the excitation forces resulted in a value of about 18 KN/mm as indicated by the horizontal line in Fig. 15. This shows that the bearing modifications using the original rotor were not qualitatively sufficient.

~~202 A distinct improvement of the overall oscillation behaviour becomes evident only after having decrease bearine dth g distanc rotore th n .eo~~

heliudesigw e th ne r mf no Fo turbines s recommendei t i , d tha ensuro t t quieea t runnind gan stable behaviou firse th r t critical frequency shoul higs a possibls he a d b thad ean t turbulence straightening sheets arrange shafe th tn d i seal s shoul providede db additionn I . , non-sensitive bearings such as tilting segment bearings, should be used whereever possible.

~~Non-Achievemen Nominae th f o t l Power~~

Withi e firsnth t three year maximua s m electrica e exceede coulb W t lM dno powe 8 d2 f o r becaus seal-gap-excitee th f eo d oscillations bearine th , g problems shafe th , t unbalancee th d san thermal distortions. After having modified the rotor of the high pressure stage, nominal values of pressure, temperature and could be reached, but nominal power could not be reached. In order to understand the reasons for this power deficit, measurements were taken at various steady-state operating levels. Of particular interest was that level of operation where the actual operating values largely corresponded wit nominae hth l design parameters.

~~No. 1 2 3 4 5 6 7 8 9 10 11 Unit B/D 0.72 0.72 0.72 0.72 0.5 0.5 0,5 0.5 0.5 0.3 0.5 -~~

~~SV - - 2.75 3.5 4.0 4.0 4,5 4.5 4,5 - - -~~

~~V 1,6 1,6 1.7 1.6 1.6 1.6 %0~~

~~Per 0,68 — — - 10.12 15.40 18.67 21.82 47.95 31.63 23.57 kN/mm B/D Ratio bearing width / shaft diameter SV Ratio bearin / shaf p gga t diameter f Minimum relative bearing gap~~

~~50r~~

~~qer 25" X X~~

1 ——— 10——— 1 — ——— 1——— 1 ——— 1— 1—— — ————4 — ————1 1 ————— —— 1 1 1 0 1 9 8 7 6 5 4 3 2 Nr 1 . 2 - Bearing 1 s consistin multiplf go e plain surfaces with/without groove 9 3- Wedge-type bearings 10-11 5 tilting segment bearings

~~Fig. 15: Permissible gap excitation coefficient qer for the different applied bearing types resp. bearing geometries~~

203 At these conditions the nominal speed, the helium-inventory, temperatures at the inlets into the compressors and the HP turbine, and the inlet pressure into the compressor corresponded with the design. The plant should have reached its nominal design power, but the actual electrical power of the plant amounted to 30.7 MW (nominal: 50 MW) and the heating power reached (nominal66.W 5M : 53.5 MW).

accuratn A e analysi showns sha , thadeficience th t powef yo r productiodeficiencieo t e du s nwa s inside the turbo-set. The blading efficiencies did not reach the nominal design values and the helium-mass flo we e sealincoolin th th rate s (whicr d ga gfo s gan e deductear h d aftee th r compression beed larger ha )n fa r then predicted. Sinc loss emi t helium-flow doe t producsno e powee turbineth n i accordingln r a , y large power loss resulted. Additionally, large losses occure insufficieno t e ddu t flow guidance fro inlee mth t diffusor intfirse oth t blade row.

Finally, EVO had decided to accept the helium turbine from the manufacturer in spite of non- conformance wite supplth h y contract. This decisio s madwa n e because helium-specific experiences could nevertheles gamee sb d wit actuae hth l plant condition becausd san plane eth t could be operated (in spite of the lower electrical power) as a cogeneration plant satisfying the required district heat demand.

Design Modification Reduco t s Powee eth r Deficit: observee th s A d power defici beed ha tn explaine vera o dt y large extent, several measures were identifie enablo dt reductions eit measure On . e woul optimizo t e db floe eth w conditione th r sfo inflo d outflowan w e compressoareath f o s turbind an r e section orden i s minimizo t r e eth pressure drop achievlosseo t d optimusn an e a m inflow int bladinge oth . Moreover reductioa , n of the blade gap losses would be required. That could be achieved by a reduction in rotor vibration and better selection of materials. The materials for both the rotor and the stationary blade carrier shoul selectee db optimizo dt e thermal expansion achievo t s e mmimum gapt a s operating conditions e approacOn . h replacemene woulth e db non-coolee th f o t d austenitic stationary blade carrier in favor of a ferritic one, with cooling provided at necessary locations. Preferabl shoulC ° coolin o provideyn 0 e d b 75 t ga allt da , taking into account possible improvements in avaible blade and rotor materials. A third approach that seems to be possible would furthea e b r optimizatio stationare th f no y blade profile rotod an s r blade profiles.

e modificationTh s require achievo dt nominae eth l currene poweth r fo rt design would have required such large expenses desig,w thane na t would have been economically preferable. Therefor concep turbosew e ne th r ea beed tfo ha tn prepared shows essentiae a , Th Fig. n i 16 .l difference in this new design is that the turbine is no longer split up into high pressure and low pressure sections. Moreover bot hige hth h pressurpressurw lo d eean section compressoe th f so r have been optimized. A larger number of blade rows, compared to the former design, because s selectebladine ath wa reactior % dfo g 0 insteadesign% n10 l turbo 0 f rati5 Al do f . o - machines run at a speed of 90 Hz, so that a gear must be provided between turboset and generator. If this concept is applied the nominal power could be reached at the previously provided circuit parameters.

~~Rotor Seals~~

The sealing systems of the rotors at the penetrations through the housings are a good example numeroue foth r s systems which have prove excellenn na t functionality sinc commissionine eth g facilityO EV e o.th f

~~204 LP compressor HP compressor 15 stages 26 stages~~

~~1430 1930~~

~~Generator~~

~~Revolutions 90 s-1 -——- 50 s-1~~

N> Fig. 16: New concept of the 50 MW helium turbine S! Fig. 17 shows the schematic design which seals the main circuit against the bearing oil circuit, and the bearing oil circuit against the atmosphere.

Sealing against the atmosphere occurs by means of a sealing oil type gasket. One half of the introduced sealing oil flows together with the discharged bearing oil from chamber (6) into a close pressurw dlo tankl othee eoi Th sealin.e r halth f getl fo g oi s into contact wit ambiene hth t atmospher returnd ean s through pip ) directle(8 y intsealine oth l tankgoi .

For the orderly functioning of the system it is most important that the pressure differentials indicated are maintained. Since the power of a closed gas turbine cycle is regulated by means of a pressure change in the main circuit, it is necessary to regulate the pressures in the auxiliary circuits accordingl orden yi r ensur appropriate eth e sealing function providee Th . d systes mha proven its orderly function without any operational problems.

The last stage of the HP turbine can be seen on the left side of the figure. Between that stage and the bearing three labyrinth seals, separated by the inner chamber (2) and the outer chamber (3), are provided. Pure helium at high pressure is introduced with a defined overpressure into chamber (2). From ther lefe t flowei th t o intt maise oth n circuie righte th th n o wels i t ,a t s a l direction of the bearing. From the outer chamber (3) one portion of the sealing helium is directe correspondine th o dt turbinegP L systeothe e e Th th .rt m a portio f heliuno m flowo t s the right throug hthira d labyrinth seathered an l , together wit bearine parhth a f o t g oil, through pipe (4) to a helium-oil-separation stage, where both media are returned to the corresponding circuits.

~~Summary~~

At the design and construction stage of the EVO plant new and still insufficiently known technologie appliede b o t d . sha Nevertheless this prototype plant demonstrated, after having settled initial difficulties, that suc heliuha m continouturbina n i n ereliabld ru plan san n ca te operation.

~~3.1.1.B Shortened e SummarNecessarth f o y y Improvements s a propose, y b d Siemens/Kraftwerksunion (KWU)~~

(Ref. [21], [29]) operationae Th l experiences wit turbosee hth resultpowee e wels th th a t s f a lso r measurements EVO-turbomachinere oth f summarizee yar d below. Base thesn do e results improvementd san modifications have been propose Siemens/KWy db U (see Fig. 16).

~~Operational Experiences~~

Self-excited shaft bending vibrations, induce seal-gay db p excitation HP-turbine th t sa e rotod ran amplified due to an unfavorable bearing design, resulted in a rubbing within the labyrinth seals radiae th an t da l blad caused ean d blade defects radiae enlargede Th b . lo t gap d sha , leading tolarga e increas coolind sealine an th s lossess f eo gga gga , that wer t considereeno e th n di original thermodynamical design.

Subsequently bearine th , ordegn i axiao m t r m rotolP reduces 0 distancH wa r60 e th y db t ea improv runnine eth g behaviou rotore th additionn I f .o r slide th , e plain bearing desigs nwa modified in order further to stabilize the rotor.

~~206 Legend:~~

~~1 Sealing heliu HP-circuit-heliud man m~~

~~2 Sealing helium inlet~~

~~3 Towards LP - sealing helium circuit~~

~~4 Sealing heliu bearind man l outlegoi t~~

~~5 Bearin l inlegoi t~~

~~6 Bearing oil and sealing oil outlet~~

~~7 Sealin inlel goi t~~

~~8 Sealing oil outlet~~

~~567 8~~

~~K) Fig. 17: Sealing helium and sealing oil systems at the HP-turbine 3 2. Power Measurements~~

After having carried out the above described modifications, there still were large deficits of the turboset power generation compared to the original design. Whereas the nominal power was 50 MW, the actual maximum power was 30.5 MW. The nominal design heating power of 53.5 MW was exceeded however and amounted to 66.5 MW. Power measurements and recalulations were carriethrer fo t e ddifferenou t operating conditions nominae th r Fo l. operating conditions, confirmes iwa t d thaappliee computeth U t dKW r code same camth eo enominat e resulth r fo t l power of 50 MW, when using the same assumptions for the thermohydraulic and mechanical desigturbomachinere th f no circuid yan t thaformerlyd ha t been use Guty db e Hoffnungshutte (GHH). When insertin latese gth t measured actual parameter circuite turboseth e f th o f ,sd o an t e measureth d electrical powe t nominaa f 30.o rW 5M l condition s recalculatewa s e b o t d 30.7 MW, which is nearly identical.

~~The following power losses have been measured and recalculated for the 100 % nominal circuit conditions:~~

~~Turbomachinery:~~

~~W M 3 1. • Flow inlee losse e th tth n i diffusosn i d an r blading of the LP compressor~~

~~W M 0 4. • Flow inlelossee e th tth n i sdiffuso n i d an r blading of the HP compressor~~

~~• Blade gap losses, flow losses in the HP-turbine 3.9 MW~~

~~• Profile losses due to remachined blades 2.4 MW after having detected damaged blades in the LP turbine~~

~~• Increased sealing and cooling gas flow 5.3 MW rate in all four machines~~

~~Total Turbomachinery Loss 16.9 MW~~

~~Additional loss due to higher pressure 2.6 MW drop in the circuit~~

~~The total power loss in the circuit was 19.5 MW .~~

This means that in order to overcome the power deficit, it was necessary to modify the turbomachiner s componentit d an ye mai th s wel a sns a lcircui y thae n sucth wa i tt a h theoretically assumed design parameters would be fulfilled by the mechanical design of the turbomachiner circuite th f o .d yan

~~Flo Bladind wan g LosseTurbomachinee th n i s s~~

~~In order to overcome the power losses due to flow and blading losses, the following countermeasures could have been taken:~~

~~208 Modification of the inflow and outflow sections of the compressors and turbines (experimental tests woul required)e db .~~

Reduction of the blade gap losses by designing essentially vibration-free rotors. The stationare material th roto e d th an r ysfo blade carrier shoul relatee db eaco dt h other in such a way that minimum gaps will be achieved at operating conditions ( e. g., replacemen non-coolede th f o t , austenitic stationary blade carrie ferritia y b r c one). Cooling must be provided as needed when the ferritic material requires lower operation temperatures.

~~Optimizatio appliee th f no d profilestationare th r fo s y roto e bladeth rd bladesan y sb using a reaction ratio of 50 % at both compressors instead of 100 %.~~

~~Sealin Coolind gan Losses gGa s~~

excessive Duth o et e vibration rotorse th f so , labyrint e partth f so removee b ho t seal d y db sha machining, thus increasing their clearances. This resulted in the additional sealing and cooling gas losses. amountinMW 3 5. o gt

~~Summary~~

~~Proposed improvements:~~

• By designing nearly vibration-free rotors, the gaps required in the labyrinth seals could reducede b . Rotor improvements could have been achieve reductioy db rotoe th f rno weights, provision of a single shaft turboset (see Fig. 16), reduction of the bearing distance, and the use of bearings with a degree of high damping. requiree Th d coolin • flos gga w rates shoul minimizede db currentle Th . y available technolog e coolinth r f rotoryfo go bladed e an availabilits s th wela ss a l f higyo h temperature resistant materials permit reductio cooline th f flos no gga w rates.

~~3.1.1.C Generatio f Acoustio n c Emmission e Turbomachinerth y b s d Propagatioan y n into Circuit (EVO)~~

~~(Ref. [30], [31], [32], [33])~~

planO locates ti EV e Th d centecite close th yth f o eOberhausent ro . Therefor necessars wa t ei y to ensure a nearly noise-free plant operation. The operator measured a noise of 70 - 80 dB directly in the area around the turboset and outside of the machine hall of < 50 dB.

l turbomachineal Likr fo e e largesth s t e portioacoustith f o nO c EV emission e th f o s turbomachineries is propagated, even against the helium flow direction, into connected piping, where it can excitate vibrations and thus develop mechanical loads on portions of the piping, thereby possibly leadin fatiguo gt e fractures orden I .investigat o t r e these problems numerous measurements were carried out. In this report only one typical example, namely the coaxial gas duct between recuperator and HP turbine, is described. Figure 18 shows the schematic arrangement of this piping including the flow baffles and the measuring points.

~~209 leve4 G l to heater from heater~~

~~microphone with probe pipe and protection pipe~~

~~from heat exchanger~~

~~HP-turbino t e~~

~~Level G4~~

~~81'5CD~~

~~Fig. 18: Koaxial gas duct between heater and HP-turbine~~

~~210 Figshow9 1 .noise sth e spectrum leve measurint a l nominat a 3 gG l power parameter totaa t sa l sound power level of 140 dB emmitted from the turbomachinery.~~

The measurements have shown that the overall level of the acoustic emissions rises with increasing electric power. At nominal plant parameters the maximum sound level within the piping section . Fig investigateshow0 dB 2 .8 tota e sth 14 ls soundwa d power level, averaged over the level circumference of measuring point G3 depends on the power level.

measuree Th d frequency valueaveragee th f so d rotary sound powemaximue th t a r m achieved electrical powe 377s . Figrshowwa 1 0 Hz ) 2 (=3 .rotare th s MW 0y sound power averaged ove circumference th r dependans ea electrie th n o tc power level (measuring point G3). These measurements hav broaea d rang f scatteeo t generallbu r y increasing wit generatoe hth r power. Small changes in operating conditions result in large changes of the sound propagation, so that at the same measurement position large fluctuations of the rotary sound power at the same modal composition of the sound field have been measured.

~~B d 0 14 + 0~~

~~-10 -~~

~~-20 -~~

~~-30 -~~

~~-40 -~~

~~-50 -~~

~~-60 -~~

~~-70 -~~

~~I I I i I I I -80 I [ 0 100 200 300 400 500 600 700 800 900 1000 10+01Hz~~

~~Fig. 19: Noise level spectrum at measuring point G3 at nominal power parameters~~

~~211 150 Measuring point G3~~

~~CO 148 I ~ 146 § Q.~~

~~"c 144~~

~~142~~

~~140 0 4 5 3 0 3 5 2 0 2 5 1 0 1 Electrical power [MW]~~

~~Fig. 20: Total sound power level averaged over the circumference dependancn i powee th n ero level~~

~~146 Measuring poin3 tG~~

~~Ef 144 2,~~

~~|> 142~~

~~--' O o 1 8.140 •o D 0O3 131080 s^ DC 136~~

~~134 10 15 20 25 30 35 40 Electrical power [MW]~~

~~Fig. 21: Rotary sound power level (averaged over the circumference) in dependance on electric power level~~

212 Knowledg modae th f eo l compositio soune th f dno field enable calculatioe sth spatiae th f no l distribution of the sound pressure, which is the cause for the piping vibrations. It has been observed, that large differences occurre circumferentiae th n di l pressure level distribution, with observed peaks at the bottom. The measured values at measuring point G3 showed a good coincidence between all measurements and the precalculated values for the sound pressures.The effect of acoustic emmissions and their propagation into the piping have been investigated regarding their influence on the mechanical loads induced by excitated vibrations. These investigations have shown that a fatigue failure of the liner guiding the flow between the fossil- fired heater and HP-turbine is not credible, because the largest stresses have been determined approximatele tob N/mm5 y2 2.

~~A fundamental fact thas beeha t n foun s thai d t oscillation amplitude e largee lowsar th n -i r frequency range than those measure rotare th t dya tone turbine P frequencH e th .f yo~~

~~Turbomachiner- V 3.1.HH 2 y~~

~~(Ref. [23], [24], [34], [35], [36])~~

~~The HHV facility was operated by a consortium of German utilities, participating in the HHT- project.~~

~~Main Problems at the Commissioning~~

~~The main difficulties occurring at the commissioning in 1979 were:~~

~~Oil Ingresses~~

Oil ingress into the main helium circuit occurred twice from the turboset seals. The first ingress, which amounted to between 600 to 1200 kg of oil, was due to a serious operator error during the commissioning phase. The HHV-plant was shut down over the weekend, but one auxiliary oil pump was unfortunately not switched off, transporting this oil quantity. This accident made evident tha interlocn a t detectoa d kan r were needed e seconTh l ingres. doi s event occurremechanicaa o t e ddu l defec sealina f o t g element. However than i , t cas ingressee eth d oil quantity was negligibly small and was immediately indicated by the detector. At the first incident the ingressed oil was partly coked and formed thick deposits on the cold and hot surface turbomachinere th f tes e o scircuite th t th n i f ducts o d ga ,d especiallt san y an ho e th n yi section. The effected surfaces had to be cleaned mechanically, by baking-out (partial evaporatio cokinr no chemicall y oil)f gb o r ,o y cracking wit additioe hth hydrogef no r otheno r additives secone th t A d. incident quantite th s a ,f ingresse yo l quantitdoi vers ywa y smallt i , was removed by cracking at 600 °C (with the use of additives).

Excessive Helium Leakrate After having modifie maie dth n turbine flange joint (see chapter 2.2.4 pressure )th lead ean k tests of the HHV at ambient temperature showed a good leak tightness for the flange joints of the turboset and of the main and auxiliary helium circuits. But at operating conditions (850 °C) large helium leaks were detected comprehensivd ,an e wor countermeasured kan takene b o t .d sha A first measure was to weld the lip seals provided at the flange joints of the mam circuits. Later a large leak was detected at the front flanges of the turboset, caused by a non-uniform temperature distribution during operation resulting in thermal stresses forming local gaps of about

213 1/10 mm. To overcome this problem, the cooling gas distribution and flow rates within the turboset housing were modifie improved temperature dan th d dan e distributio s improvenwa d sufficientl thao locae s y th t l gaps were prevented course commissionine th th f n I eo . g runs another large leak occurre maie th nt da flang eturbosee jointh f o tt t no seal(whep d li e sha nth been welded). The outer sealing ring was partly ruptured due to operator's overpressurization of the purge groove. Modifications to the interlock system were then made.

~~Trial Run: Demonstratio f Safno e Quick Shutdown~~

After having overcome the leaks of oil and helium, and the associated cooling problems, the stepwiss wa V e HH brough fulo t t l operatin bar1 5 d . gDurinan h conditionC 0 ° 6 g0 a 85 t sa trial run the functioning of the instrumentation, control and safety systems and the general operating function of the HHV were demonstrated. A principle task was to demonstrate the emergency shutdown and the reliability of countermeasures in case of incidents. For this demonstratio e e turbosecoolinth th n s d compressoga gan t d beeha rn switche t fula f l of d operating conditions turbosee Th . t mus slowee b t d down electricall fulya o withilt stos 0 pn9 in order to prevent an unpermissible heatup of the circuit by the rotation energy of the rotor shaft. Subsequently, the HHV was returned to the full operating condition. This safety demonstratio turbomachinere th r nfo shows yi Fign n. i 22 .

~~Overall Operational Performance~~

~~After having overcome its initial problems, the HHV was sucessfully operated for about 1100 hours,of whic turbomachinere hth y operate. °C r abou 0 hourdfo 5 85 32 tt sa~~

~~——•— Hot-gas temperature T MW — Cooling gas temperature .K — Drive power P System pressurep~~

~~40--40 500~~

~~Intentional shutdown \ 10-- TurbomachineryI in operation~~

~~Cooling compressorgas operationin 0-L 12 1U.1M1 Time~~

~~Fig. 22: Startup of the HHV from cold condition and after a quick shutdown (hot condition)~~

214 turbomachinerV DurinHH e e desigth th g f o n y comprehensive precalculations were made regarding the operational behaviour at various partial loads and at nominal load. These precalculations were be counterchecked during the experimental operation of the HHV plant for the nominal plant parameters. Comparisons betwee e precalculatenth e measureth d an d d parameters are given as follows:

~~Table 3.1.2/1: Plant Parameter Nominat sa l Load~~

Precalculated Measured Turbine inlet 49.4 bar 49.1 bar C ° 0 82 818 °C Between turbind ean 45.1 bar 44.8 bar compressor 785 °C 780 °C Behind compressor 51. 6 bar r 51.ba 9 C ° 0 85 C ° 9 84 Required motor power 45 MW W 39.M 2 mase H s flow Kg/2 21 s 211 Kg/2 Pressure drop r 2.2ba 1.r 9ba

agreemene Th t between precalculation measurementd san excellents swa . Whe folie nth o whig uncertainties hi the precalculations and the restrictions at the measurements during the operation are considered:

• mas turbomachinse floth i wh e canno measuree b t d directly • mass flows in stuffing boxes and labyrinth seals can only be estimated • coolinstatoe onl n th roto e estimateflowe s d ca ryb th g ga i an r h s d concludee b n itca d thathermodynamie tth c design data were achieve eved dan n exceededr Fo . example the compressor and the turbine had a better efficiency than assumed at the design. This was derived fro measuremene mth electricae th f o t l drive powe frod direce an mrth r indireco t t measured mass flows, pressures and temperatures at the turbine and compressor inlet and outlet

~~Dynamical Behaviour~~

Extensive measurements were made during the commissioning phase as well as during the 60 h trial run with regard to the shaft rotor oscillations of the turboset with the measured results compared wit precalculatee hth d values. Thu firse sth t resonance frequenc calculates ywa o dt be approximately 1700 rpm, whereas the measurement showed, that it is hi the range between 200d 170an 00 rpm secone Th . d resonance frequenc precalculates wa y r abovfa e e eth b o dt nominal speed, whic confirmes hwa operationy db .

rotoe Th r shafrunnins wa t g quietl l operatinal t ya g conditions e runninTh . g behavious wa r better than preplanned wit shafe hth t oscillation amplitude t nominasa l ordespeee 0 4 th f i do rh to 60 /mi. This is usual for large steam turbines. A detailed comparison was carried out for

215 three startup and shutoff runs as well as for the 60 h trial run. The measurements and comparisons concerned rotation-frequency, double-rotation-frequenc self-excitated an y d vibrations. The coincidence of measurements and calculations was good. The measurements confirmed that rotoe th r shafsatisfactorils wa t y balance yielded dan d valuable information regardin rotoe gth r shaft behaviour at the startup and at the resonance frequencies. However, it should be pointed out that the good agreement between measurements and precalculations most probably occurred, becaus rotoe eth r shafonld ha yt slight imbalance onld san y slight thermal distortions. Other concern precalculatione th f so s wer bearine eth g properties. More details abou dynamicae th t l properties and behaviour are shown hi the following figures.

Figshow3 2 mas.e th s s rotodistributionV rHH shafe indicated th an tf o s measuremen e th s t points. Figshow4 2 .dynamica e sth l behaviou maie th nf ro roto rrunupe shafth t a t , measured at point A. The figure shows clearly, that the rotor shaft was excellently balanced. Fig. 25 show dynamicae sth l behaviou maie th nf o rroto r shaf t rundowna t , also measure t poinda . A t The figure indicates, that ther beed eha n some slight thermal distortion rotoe th f rs o shaft . Fig. 24 and 25 show moreover that the first resonance frequency is at 1700 - 2000 rpm.

~~Required Modificatio f Instrumentatioo n n Monitorin Temperature gth Rotoe th f reo~~

~~The continous temperature measurement provided at the rotor, which consisted of a slip-ring- transmitter replace s t provreliable b no wa telemetri a o d ed t y di d,eb an c measurement.~~

~~Cooling of Turbomachinery~~

Rotor, stator, blade inle feee outled wels th a tan t s a l t nozzel turbomachine th f so e were cooled by mean f colso d helium conducted through cooling channels. Satisfactory rotor coolins gwa especially importan e safth e r operatiofo t plante e newlth Th f n.o y designed rotor cooling proved to be very effective, assuring that the disk temperatures could easily be kept below 400 °C. The measured coolant gas flow rates were 10 to 20 % larger than the planned flow rates, so that the actual temperatures of the rotor disks and blade feet were distinctly lower than forecasted. Only at the turbine inlet (where the cooling helium temperatures are the highest), were the measured values 20 - 30 °C higher than estimated by the design, but distinctly below permissible values. Fig show6 2 .pressur e sth temperaturd ean e pattern bladine th n si g channel cooline th an n di g channel rotoe wels th a srs a l dis bladd kan e feet temperaturese b y ma t I . noted that in addition: nozzlee th functionins r cooline fo statose wa d th Th r an r gfo g temperaturwelle • Th . e inlee ath t t thanozzlC ° n5 precalculatede3 flangeo t lowes 0 3 wa sy rb .

~~• The only modification of the cooling system that was required was for the front flanges turbosee oth f t (equilizatio coolinge th f no i ordeh , preveno t r t flange splitting).~~

~~cooline Th g syste • m functione wello ds , tha operatinn a t g temperaturr fo f 100eo C ° 0 the turbine seems to be possible.~~

~~Shaft Seals: Helium-Tightness~~

~~Shaft seals generally performed well as follows:~~

~~3-circuie Th t floatin • g ring sea rotoe l th use r dshaffo t seal, whic beed hha n pretested ia nspecia l test stand, demonstrated adequate behaviour under helium operating~~

~~216 LIZ~~

~~Equivalent diameter [mm]~~

~~8 a I o- I~~

~~(0 B> N) 00 RMPX (MM-31~~

~~2500.0~~

~~5000.0~~

~~] Frequencn mi / 1 [ y~~

~~Measurement point A~~

~~Fig. 24: Dynamical behaviour of the main HHV rotor shaft at the runup RMRX IMM-31~~

~~2500.0~~

~~5000.0 7.5 Time [min]~~

~~7500.0~~

~~Frequency [ 1 / min ] .0~~

~~Measurement point A~~

~~Fig: Dynamica.25 l behaviou rotomaie V th rnf rundowHH rshaf o e th t ta n 400- Material temperatures: Rotor disks~~

~~350-~~

~~300-~~

~~Material temperatures: Blade feed tan interim pieces I Hot-gas temperature in turboset blading channel 850^~~

~~800-~~

~~750-~~

~~300- Rotor cooling gas temperature~~

~~2501 bar 54- Rotor cooling gas pressures 52- 50- Hot-gas pressuresn i 48- blading channel 46- 44-~~

~~Fig: Temperature.26 pressure-patternd -an bladinsn i g channel, cooling channels rotod ,an r disks~~

220 bar1 5 d . an Mino C condition° r0 modification85 f o s beed ha sn requiree th r fo d regulation of the sealing oil system in order to eliminate the disturbing effect of gas bubble releasesoperatioV HH n total I e . nth , showed thae selecteth t d shaft seal fulfilled expectations observee Th . d rotor shaft oscillation t hav effecy no ean n d o tsdi the sealing behaviour.

~~• The static seal at the rotor shaft also functioned without any problems.~~

• After performing the above modifications, the sealing and cooling gas system including the cooling helium compressor, functioned well. The purification system for the circulating sealing gas did not function satisfactorily, however. The sealing gas flowing from the labyrinth seals to the bearing pedestal chamber becomes loaded with oil aerosols (« 50 grams per hour oil at steadystate operation) and must be cleaned by means of a separater and filter system in order to reach the required cleanliness of the gas before it is returned into the circuit. This oil separation system, consisting of a cyclone separator and a wire-mesh and down stream fibre filters, needed further improvement l ingresoi n A s. int innee oth r helium longe o circuin s r wa tobserve d after performin above gth e modifications.

• After performin above gth e described modification horizontae th seae f th so l l turboset flang frone s welth a e shafe s t a lth sea f t o lsea l carrier proved their reliabilitye Th . measured leak tightnes f theso s e flange joint bettes swa r than 10"3 r secondTorpe 1 r . The weld lip seals of the hot gas duct were tight. Thus the helium losses against the ambient environment amounted to 10 - 20 NmVd at 51 bar (out of total helium inventor f abouyo t 8000 Nm3).

~~Generation of Acoustic Emissions bv the Turbomachinerv and Propagation into the Circuit and Environment~~

The noise level specified outside the HHV concrete building was less than 50 dB. The actual noise leve HHV-turbomachinere th l insidd yha buildine eth closn g(i e vicinit turbosete th f yo ) (estimatew lo wa o s d approximatel tha) furtheo tn dB 0 6 r - measurement0 y5 s were necessary.

During the trial run measurements of the sound power spectrum were taken at four different measuring point connectioe s th (tw t a ocoaxiae th t na nozzlel o e turbosetestw th tf d o s an t section helium-circuite th f )o . Thidons determinswa o et spectrue eintensite th th e d th m f an y o noise generated and propagated by the turbomachinery and to investigate its influence on resulting vibration wale insulatiod th an lf so n element overale th f so l circuit (see Fig. 27).

The frequency spectrum of the analysed microphone signals consists of a stochastic, broad spectru f noismo e contribution e narroth f o w d spectruan s f rotatiomo n frequenciee th f o s turbomachinery compressoe Th . r section contribute fundamentasa l frequencd 280f an yo z 0H the turbine sectio nfundamentaa l frequenc averagee Th 420f yo . 0dHz total sound power within helium-circuie th pipins it d gan t increases wit powedrive e hth th f e maximuo r a moto o t p ru m level of 160 dB. The total sound power as dependant on the drive power at measuring point 13 is shown in Fig 28.

The modal composition of the acoustic emissions released into the hot gas duct was determined also. Fig. 29 shows the modal compositions measured at different pressures and temperatures during the trial run.

~~221 .0130dB + 20 dB Compressor outlet~~

~~-10 -~~

~~-20 -~~

~~-30 -~~

~~-40 -~~

~~-SO - •eo -~~

~~-70 -~~

~~-so 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 10*" Hz~~

~~.0130d + 10d8 B Turbine Inlet~~

~~0 10075 0 0 12550 0 0 15025 0 1750 0 2000 2250 2500 z 10H *"~~

~~B .0150dd 0 2 + B Coaxial test section~~

~~-10 -~~

~~-20 -~~

~~-30 -~~

~~-40 -~~

~~-50 -~~

~~-60 -~~

~~-70 -~~

~~-80 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 z 10H * m~~

~~Fig : Soun.27 d power spectru varioumt a s location HHV-plane th f so t~~

~~222 162~~

~~0 16 , — CO o "o 'o > 158 .0) I OD. 156 TJ~~

~~O 154 IV) *~ 152~~

~~150 5 2 10 0 2 15 30 35 40 Electrical power [MW]~~

~~: TotaFig28 . l sound power leve dependencn i l drive th en epoweo r~~

~~160~~

~~60 22-May-198 o hourt - s20 trian ru l 1~~

~~150~~

~~140 m •o~~

~~130~~

~~120~~

~~20-May-1981 21-May-1981 22-May-1981 861 °C 49.1 bar 862 °C 49.6 bar 48.C ° r 9 7ba 85 110 -2 m Fig. 29: Modal composition of the acoustic emmissions into the hot-gas duct~~

223 e knowledgTh modae th f eo l compositio sounn ni d fields includin magnitude gth phasd ean e angle enable calculatioe sth totae th lf no spatia l distributio soune th f dno pressures. Usine gth developed calculation procedure, the dynamic response of the circuit insulation resulting from excitatioe th soune th y calculatedde nb b fieln dca .

~~3.2 Systems/Components/Materials~~

~~O 3.2.EV 1~~

~~(Ref. [16], [19], [20], [36])~~

~~Helium purification~~

orden I gaio t r n experience heliue th , m purification-syste fossil-firee th f mplano O s EV wa td designed in accordance with the requirements for a nuclear heated plant, except without provision for the removal of radioactive impurities. The design throughput was 100 kg/h and th substancey e requirean r floe fo wTh m .d schem pp cleanlinessystee 1 th f < meo s i s wa s given in Fig. 30.

The inflowing helium containing impurities was conducted through a dust filter (F) into the gel- filled molecular sieve T (adsorption of H2O and CO2). Subsequently, it flowed through a recuperator, s coolewherwa t di e dow, inte liquiK oth abouo ° nt d0 9 tN 21iqu-cooled heat exchanger where it was cooled down further to 80 ° K and where N2 and O2 were partially removed. From ther flowet ei separatoa o dt r (separatio gaseouf int) n2 o e oO th liquid r o san 2 dN low temperature absorbers (GA), which were filled with several types of gels for the removal of residual N2 and O2 as well as of other gaseous impurities.

Durin firse gth t operating phas requiree eth expected dan d cleanlines heliue th f mso coult dno be reached even with a continously operating purification system. The cause was found in the sealing oil circuit, since at various locations the sealing oil contacted and dissolved air. This combinatio dissolvef sealind no an int y mail e foun- ogoi th wa dah ns dit heliu m circuite Th . proble solves m wa providin y db additionan ga l degasifie sealine th r gfo r oil.

Table 3.2.1/1 show impuritiee sth s measure heliue th i dmh maiplantO e nTh EV . circui e th f o t first line show valuee sth s without degasificatio secone th d dnan line those with degasification sealine ofth g thiroile Th .d line show contene sth impuritief to maie th nn si circui heliue th f ti m purification system is not operated. Lines 4 and 5 permit a comparison with AVR and Dragon (except for radioactive impurities). Line 6 gives the maximum permissible values as specified for HHT.

~~The overall performanc heliue th f mo e purification syste judges mwa d satisfactory. Sinco en oxidation bed for the removal of H was provided, no statement on the removal of that type of~~

~~impurit madee b n .yca 2~~

~~Forces from Gas Ducts and Heat Exchangers on the Turbomachinerv~~

suspectes 197 y initiae A wa plan th tO Ma t 5 i n li tEV d startu e thath f tp o larg e unplanned dan undefined forces from the ducting and from the heat exchangers were acting on the turbomachinery. These large forces had an adverse effect on the running behaviour of the turbomachinery and moreover these large forces acted adversely on the horizontal flange joints of the turbomachinery, affectin heliue th g m leak tightnes locao t e l sdu splittin flange th f geo joint.

~~224 Legend: K = Compressor = Dus F t Filter T = Mol-siere WT = Recuperator V = Heat Exchanger = Absorbe A r GA = Deep Temperatur Absorber VP = Vaccuum Pump~~

~~Turbine circuit with heat exchangers~~

~~N> N> Fig. 30: Helium purification system for EVO Table 3.2.1/1: Impuritie Heliun si m Circuits~~

H20 CO2 H2 CO CH4 N2 02 3 m cm3/ withouO EV t degasification < 0.1 20 2 2 < 0.5 240 20 of sealing oil EVO with degasification of 15 1.4 5.5 < 0.5 8 < 0.1 sealinl goi EVO after a 2-week operation of the main circuit 15 22.5 22 13 177 1 without He-purification operation AVR Juelich * 0.1 0.5 30.. 50 125 0.4 20 Dragon * 0.05 0.02 0.5 0.4 0.05 0.6 Specified impurity limitf so 5 1 50 50 5 5 0 HHT*

~~Radioactive impurities are not included~~

~~Operational measurements showed thaturbomachinere th t y foundation move t startuda p firsn i t directioe on whed nan n reachin operatioe gth n temperatur opposite th n ei e direction.~~

Subsequent inspections, which included disconnection and reconnection of the ducting to the LP maie th nturbinf o hea d tan e exchange ductine th o t r g showed large deviation e initiath f o ls alignment of the rotor axis of the LP turbine, which initially was accurately aligned. A realignmen s requiredwa t t becamI . e evident thadesige th t n provide r compensatindfo e gth thermal expansion torsiond ductinge an s th f o s , heat exchanger turbomachinerd an s t no s ywa satisfactory i.e. the original provisions for slide bearings, support springs and bellows were not sufficient.

~~Under the restriction that the very compact plant arrangement could only be changed slightly, successful additional provisions were taken as follows:~~

~~• Provisio additionan a f no l belloduce th tw n betweei heae nth - t exchangeLP e th d an r turbine~~

~~• Additional insulatio foundatioe th f no n steel structure~~

~~• Modification of the forces of some support springs~~

~~After having performed these improvements no further severe problems or misalignments were observed.~~

~~226 Ducts HoGa t s~~

heliut ho plant AO e m coaxia) Th EV (75. °C e ducts 0 th ga lf t sdesigo ho selectes e nwa th r dfo flows in an inner liner tube, with a fibre-type insulation provided on the exterior, enclosed by a inner pressure bearing tubecoldee Th . r helium flow surrouna n i s d annulus, wit outee hth r tube designe r fuldfo l system pressure.

~~This desig alss onwa applie bellowe th flango dd t san e joints l flangAl . e joint providee sar d with wel sealsp dli .~~

~~The design chosen for the hot gas insulation proved reliable.~~

~~Corrosion. Erosio Turbinn no Compressod ean r Blades~~

No signs of any corrosion or erosion of the turbine and compressor blades or other helium- components coul observee db inspectionse th t da , modification planO EV t repaird e an th s f so and EVO turbomachinery.

~~3.2.2 HHV~~

~~(Ref. [23], [24], [34]) ] 7 3 ,[ [35] , ] 6 3 ,[~~

~~Helium purification~~

~~Description of Design~~

~~show1 3 Fig floe . sth w schem continousle th f eo y operating helium purification systee th f mo HHV-plant.~~

At normal operation a partial flow of 1000 Nm3/h was extracted from the test plant (total inventory: about. 8000 NmVh) and processed in the helium purification system. The extraction heliue sidth P f H em occuro e compressoth t sa r throug hdusa t filter that follow oxidatioe th s n bed. This oxidation bed contains two beds, one consisting of copper and copper-oxide on carrier containe2 O d an impurities da O C materialCO d , 2 convertee an H 2ar s . thiO d n 2 I be s.H o dt Before being conducted into the cryogenic section of the purification plant, the helium is cooled and then conducted through an activated carbon filter provided for absorbing carried-over oil and a dust filter provided for retaining activated carbon dust or other dust.

Insid e removee cryogeni ar th e counter-flo 2 th i CO h d cd sectioan wO 2 cooler/freezernH . Insid low-temperature eth absorbee ar 4 activatey CH db d e an absorbe r dA carbon , 2 e N r Th . purified heliu mthes i n returned throug hdusa t filter int maie oth n circuit.

The effectiveness of the helium purification system is continously monitored by a gas chromotograp humidita d han y measuring device specifiee Th . d cleanlines thif so s system was: l oi d an m vp 5 eac2 0. N h< d an O 2 vmp4 eac1 H , h< CH d , Coan r Co 2 A , , vpm2 5 0 H,< 2 < 0.1 vpm.

~~227 Dust filter~~

~~i -178'CQ Deep~~

~~HP-outlet~~

~~Counter flow cooler / / freezer Coolin s 28*CCgga F compressor~~

~~Activated N 2 liquid carbon -cooler filter~~

~~Cryogenix cbo~~

~~Fig: Heliu31 . m purification system~~

~~Operational Experiences~~

Operation showed that it was necessary to install within the main helium circuit an additional device for measuring the accrueing special gaseous impurities, that is hydrocarbons and oil vapor r thiFo s . purpos e"flame-ionisation-detectora provides wa " proved dan d successful.

A wholesa heliue th , m purification system reliablprovea e b well-functionin d o dt ean g system. Although the content of impurities in the helium to be purified was higher than specified and expected, the purification system lowered each of the gaseous impurities to values < 0.1 ppmv. This latter value is below the lower detection limit of the highly sensitive gas chromotograph used, which means that the actual function was better than specified.

~~s HoDuctGa t s with Insulation. Regulation Valves. Bellows~~

~~Hot gas Ducts~~

~~For development purposes three different types of hot gas duct sections were used and evaluated:~~

• One having an inner liner, providing flow guidance and a outside pressure-tight wall with Kaowool-insulation stuffe betweenn i d . wite hOn coaxia • l flow guidance innen A . r tube carrie heliut outen a s ho d rm annuluan s carries cold helium. The inner tube is provided on its outer surface with a Kaowool- mat- insulation and the inner wall of the pressure bearing tube with a Kaowool-mat- insulation also.

~~228 • One test section of the hot gas duct with inner metal foil insulation instead of the Kaowool insulation and an outer pressure-tight wall.~~

Fig. 32 shows the design of the hot gas duct with inner liner (made of Inconel 625), the surrounding volume stuffed Kaowool-insulation and the outer pressure tube made of ferritic steel, which is water cooled by means of welded-on semi-tubes.

Fig. 33 shows the coaxial design of the hot gas duct shown with the example of a bend. It consist innea f o sr flow guidance tube surroundine th , g Kaowoo t insulationma l cole s th ,d ga tube (where the cold helium flows in counterflow to the hot helium), the outer Kaowool-insulation and the water cooled pressure bearing tube.

~~The test s ductinsection ga three t th geho f o sdesign s were instrumentee th t orden i dge o t r necessary information.~~

~~Flow Regulation~~

Two valve flos ga wregulat o t s t througho e teseth e hth t section were provided. They havea nominaducs ga tt behinarrangee ho l ar e bypaswidte d th dth an n f 100di ho m s 0m branc d han before the bypass reentrance. There are also two smaller valves in the bypass line with a nominal width of 830 mm and also in the hot gas extraction line with a nominal width of

~~4 569 87 4 3 2 1~~

~~uoo~~

~~1 Inner liner 6 Fixture bolt 2 Perforated sheet plate 7 Stuffed insulation 3 Mesh wire 8 Outer wall 4 Interim sheet plate 9 Water cooling 5 Support element 10 Wel seap dli l~~

~~: FigLongitudina32 . l section duc s througga t t witho ha inner insulation~~

~~229 Legend:~~

~~1 Inner liner 4 Outer pressure tube 2 Inner insulation 5 Water cooling 3 Outer insulation 6 Cold-gas inlet~~

~~Fig. 33: Longitudinal section through a hot gas duct bent with coaxial flow~~

~~400 mm. The valves are not required to be completely leak-tight. The typical design of such a regulation valv shows ei Fign ni . 34 .~~

~~Bellows~~

Two bellows are provided in the main hot gas duct in order to compensate for thermal expansions in the horizontal direction. The bypass line is also provided with three bellows for compensatin horizontae gth l thermal expansions.

Operational Experiences with the Hot Gas Duct defines assessmene use wa th tria d Thh comparisor e n dan d0 th fo e ru l 6 d insulatioe an t th f no n properties of the different types of hot gas ducts. During the measurements the helium circuit condition beed sha n maintained nearly constant resulte Th summarizee . sar followss da :

~~230 Different Test Sections~~

• Tests with the hot gas duct with the inner liner, Kaowool insulation and outer pressure tube showed that only small gas flows were observed within the undisturbed axial length of the insulation. However, at the location of transition to the flange joint, where the Kaowool stuffing density is probably less homogenous, larger temperature differences inside the insulation as well as on the outer wall were measured. The average Nusselt- number coincided well with former measurements. However, the distribution of the Nusselt number over the circumference was less favorable than expected. This deficienc explainee b n stuffine yca th y db g density lowe% , 5 whicr1 thas hnwa normal, in orde implano t r instrumentatioe th t n devices.

• The coaxial hot gas test section with cold gas hi the outer annulus and provided with Kaowool-mat insulation showed a very satisfactory performance. foie th l r insulateFo d • duct larger temperature differences were expected betweee nth bottom and the top sections. The measurements showed that close to the flange connections some temperature peaks occurre coolee th n do outer wall, which were slightly above the maximum permissible peaks of 80 °C. The measured circumferential distributio Nussele th f no conclusioe t numbeth o t d rle n that natural convectioe th s nwa probable temperature reasoth r nfo e peaking, especiall vicinite e flangth th n yi f yeo joints. A satisfactory modification of the as-built design was not considered to be feasible.

~~Insulation~~

e comparisoTh f boto n h insulation types, stuffed Kaowool insulation versus metal foils insulation, led to the unanimous conclusion that the Kaowool insulation type is definitely preferable. However, further confirming tests at other pressures and temperatures are needed as well as tests for long-term behaviour.

~~Cooling~~

The outside water cooling, with the cooling water flowing through semi-tubes or cover shells, functioned very satisfactorily. Nowher surface th d educdi s temperaturga tt exceeho e th d f eo uncriticao , exceptw °C r 0 fo 6 t l spots with metal foil insulation, where local temperaturef o s about 100 °C were measured.

~~Regulation Valves~~

~~The regulation valves provided for regulation demonstrated reliable functioning.~~

~~Thermal Expansions and Flange Joints~~

e thermaTh l expansio ductins ga n t behaviougho correspondee th f o r d wit precalculatee hth d behaviour. The horizontal expansions were taken up by the bellows and the sliding supports located inside. Vertical expansions were accomodate suppory db t springs weldee Th .sea p dli l flange joints maintained their helium tightness even after multiple pressur temperaturd an e e cycling.

~~231 11 11 \ \~~

~~Legend:~~

disp kRa 1 8 Insulation sleeves 2 Rap shaft 9 Stuffed insulation 3 Axial bearing 10 Support bolt 4 Support bearing 11 Water cooling 5 Inner liner 12 Helium cooling 6 Stuffing box 13 Regulation drive 7 Outer housing

~~Fig. 34: Cross section through a hot gas regulation flap~~

~~Impact of Sound Load on Hot Gas Duct and Insulation~~

In the tests with the metal foil insulation, numerous strain gages were placed to determine the influence of the sound field on the fatigue strength of the hot gas duct. The main goal of these investigation validatioe th s calculationaa wa sf no l metho precalculato dt e randomly occuring vibrations from known local distributio d frequenciean n e sounth f do s pressurese Th . calculations measuremente showeth d dan s confirmed thafloe th tw guidance tube woult dno exceed its fatigue strength limit during the designed lifetime of 40 years. It is interesting to note thameasuree th t d structure responses sam e werth ef eprecalculateo e orde th s a r d responses. (Also, see chapter 3.1.2 on this subject)

~~Experimental Experiences with other main Components~~

~~Drive Motor. Startup Motor~~

~~The synchronous drive motor (45 MW) at 3000 rpm showed a trouble-free operational behaviour. The same applieds to the asynchronous three-phase motor (4.5 MW) for the start-up.~~

~~232 The gear provided between main shaft and startup motor, which is required only for this test plant, released some noise.~~

~~Coolin Compressos gGa r~~

Two redundant radial-type compressors were provided. They were arranged in a barrel-type housing and each had a throughput of 56.8 kg/s at an nominal inlet temperature of 236 °C with outlen a d inle. n ta an °C pressur t r 8 pressurba 25 9 f t 53.4 ea o f r eo 5 ba

At the commissioning large problems had been observed because of the ingress of oil via the labyrinth seals into the gas circuit. As described earlier, after taking adequate countermeasures, no further difficulties wit l ingreshoi r heliuso m leaks were experienced.

~~Coolers, Oil Pumps and Auxiliary Systems~~

~~All other vital component auxiliare th n i s y systems operated very satisfactorily.~~

~~Cooling Water Flow Monitoring: other instrumentation/interlocks~~

The test operation of the HHV showed that monitors were necessary in order to ensure that the cooling water flows hi each of the approximately 50 cooling water circuits. Moreover operationing experience showed the need for a hydrocarbon detector and for adequate interlockin sealine th l lubrication gi g oi n systems.

~~Behaviou f Protectivo r e Coatings~~

Pretests indicated that coatings were probably required for structural components which have sliding motions during the operation when exposed to high temperatures in pure helium. This was though necessare b o t t preveno yt t excessive frictio seizingd an n s wela , provido t l r efo disconnectabl ee pretests a resuljointsth s f A o t ., coatings used wer f chronium-caro e - bide/zirconium-oxide and provided on the surfaces either by a plasmaspray procedure or a detonation coating procedure. Coatings made of boron nitride powder were used for less critical positions, e.g., at the sliding positions of the inner hot gas duct liner.

~~All these coatings showe verda y satisfactory behaviour. However judgmeno n , longe th n -o t term behaviour can be made on the basis of the short HHV-operational history.~~

~~Corrosio Erosiod nan f Turbinno Compressod ean r Blades~~

inspectionse Ath t , corrosio y modification signo an n f dismantlinso e V th n i t HH a e r so th f go helium-circuie th sigy f erosioan no turbin e r th o t t na r compresso eo r blades were observed. e sievducs r catchinTh ga fo ett provideho ge th residua n di l particles fro installatioe mth r no particles accrueing durin operatioe gt provth necessarye b no o d et ndi .

~~However e reservatioth , n mus made operatio b V tshoro to eHH mak o t s ttha e y nth wa tean reliable statement on corrosion as well as on erosion.~~

~~233 Status of Magnetic Bearing Development~~

beginnine Ath t nineteee th f go n eighties engineerine th , g developmen f magnetio t c bearings (radia axiad an l l bearing catched an s r bearings heliue th r mfo ) heliue bloweth d man r turbine had ben initiated by ABB. A test stand had been set up in Jiilich. Test bearings had already been provided.

terminatioprojece AT th t developmene HH th te th f no reorientes wa t concentrated dan e th n do magnetic bearings for helium blowers. Due to the lack of funds, ABB terminated the developmen beginnine th t nineteee a t th t go n nineties.

However, in the meantime the test stand has been reinstalled and recommissioned at the "Hochschule fur Technik, Wirtschaft und Sozialwesen" in Zittau/Gorlitz. There the interrupted developmen r magnetifo t c bearings shal continuede b l .

~~3.3 Other Experiences~~

numbeA technicaf ro plantlV itemaddressee b HH s o st d an requestes da O EV e IAEy dth b r Afo canno e treateb t d sufficientl d completelan y y considerinb y g exclusivel e experimentayth l experiences from thes plantso etw . followinThae th trus i r t efo g reasons:

~~w peae lo th ko plantemperaturto O s tEV wa (75 • e ) th 0 °C f eo operatioe th nHHV-plane tim th • shor o f eo to s t wa t~~

~~Therefore a number of the items to be addressed like:~~

• effec impuritief o t materialn so s • performanc wead coatingf ean ro s • helium purification • erosion of blades by graphite particles • helium leakage ducts ga t s includinho g • valve regulatiod san n valves • general operation experiences with high-temperature systems

~~can be judged better by additionally considering the operational and experimental experiences I tesI tA facilitiesEV d an . K KV d an R AY e oth f~~

The respective technical experiences additionally gained from the AVR and the KVK and EVA II test facilities beeing relevant for the helium turbine technology are summarized hi Annex II. 4. Summarizing Conclusions on Technical Experiences

~~(Ref. [19], [21], [23], [24], [39])~~

~~1 4. Turbomachinery~~

Both the designs and operational experiences for helium turbines and compressors and their auxiliary equipment for EVO and HHV facilities are described in this report on the basis of publicly available information.

234 While the orginally planned scope of the experimental work of EVO and HHV could not be execute fundino t e ddu g limitation earld san y funding termination reasone criticae th ,th r fo sl problems experienced were found and corrected where economically possible. The principal countermeasures:

• EVO: the shortening of the bearing support distance for the HP turbine/HP compressor, the application of other bearing-types, the provision of an additional bellow in the hot gas ducting to prevent another thermal misalignment of the rotor axis and a splitting of the horizontal flange.

• HHV: successfull modificatio prevenreoccurrenco e nt th ) (1 tl ingres oi f eo s inte oth main turbomachiner intd maie yan oth n helium-circuit preveno t ) e excessiv(2 , th t e helium leek rate, and (3) the provision of detectors for hydrocarbons in the main helium circuit.

~~Positive experiences were achieved with both facilities after overcoming these initial deficiencies:~~

• Excellent performance of the gas and oil seals • Low helium leak rate • Excellen ductings t turbomachinere performancga th t f ho o , e th f o e y coolinge th f o , helium purification system (except the oil aerosole separation), and of the instrumentatio regulationd nan . dynamicae Th l performancturbomachinerV HH e th f eo patentls ywa O y excellentEV e th t bu , turbomachinery showe first da t insufficient dynamical behaviour. This dynamical behavious rwa improved sufficiently however, wherea powee sth r deficiturbomachinere th f o t y e coulb t dno overcome without significant rebuildin r exchanggo turbomachinerye th f eo reasone Th .r fo s all the experienced problems were well identified and completely redesigned turbomachinery was proposed to replace the first turbomachinery.

Germae Th turbins nga e expert EVOt sa , judgABBA experimentae KF e, th d Siemensan U KW /l experiences achieve accompanyine th d dan g analyses very positive unresolveablo N . e problems were identified believeds i t I . , tharesulte th texperience d san s achieved provid firmea basif si a new initiative is taken for helium turbine power conversion.

~~235 Annex I~~

~~ProjecT HH HistorGerman n i te th f yo y~~

~~(Ref. [ 23 ])~~

~~Subsequentl histore HHT-developmene yth th f yo describeds i t :~~

~~T projec HH Initiation 196i te th 8 f d Germao nwithi3r e th n Atomic Energy Development Program~~

~~Start of the feasibility investigation in 1969 for a 300 to 600 MW-demonstration plant~~

~~feasibilite th f o d y En phase: 1972~~

~~Germah 4t projec T e Inclusioth n HH n i Atomite th f no c Energy Development Program in 1972~~

~~Initiation of phase I of the HHT project by BBC/HRB/KFA in an international cooperation with General Atomic Company BBC/EIRd an A US , ; Switzerlan 197n di 2~~

~~Reference design: Block-type fuel elements; 3 x 360 MW He-turbine, integrated arrangement of the three turbosets in the prestressed concrete pressure vessel.~~

~~Decisio construco nt t HHV: 1972~~

~~Decisio referencT modifno t HH e yeth desig providiny nb single gon e He-turboset with block-typd an 124W M 0e fuel element 197n si 5~~

~~Decision to use the pebble bed core, like for the PNP project, for HHT in 1978; reference powe commerciaa r fo r l power plant: 124 witheliue W 0hon M m turboset~~

Decisio demonstratioa r nfo poweT pebblnHH d r an plancor d eW be n ei t M wit6 h67 1978; further goal : safetT AssessmenHH y e concepth f o t independany b t t experts (whic concludes hwa Novemben di r 1981).

Decision to cancel the HHT project in 1981 and to concentrate instead on a shorter term feasible HTR with the steam cycle: Thermal power 3000 MJ/s with block-type fuel elements and later with pebble bed core.

~~It shoul emphasizee db d thaHHT-projece th t always wa t s actively supporte Germay db d nan Swiss utilities.~~

~~236 AnneI xI~~

~~Relevant Technical Experiences additionall yd othe gainean rR Higd AY fro he mth Temperature Helium Test Facilities in Germany.~~

~~(Lit. [ 40 ] to [ 67 ])~~

~~General~~

helium-coolede Th , high temperature test reactoArbeitsgemeinschafe th f o r t Versuchsreaktor (AYR) (thermal power MJ/s5 4 : , electrica s operatewa ) lr approximatel powerdfo MW 5 1 : y 127,000 hour approximateld san y 45,000 hour heliua t sa m temperatur . Beside°C 0 s sit 95 f eo functio integran a s na l test reacto sphericar rfo l fuel elements, valuable experiences were gained for the typical helium systems and components.

e componenTh t test loop (KVK s providewa ) r testindfo maie gth n helium componenta f o s nuclear process heat reactor concept (PNP) at 950 °C (max. 1050 °C for short periods). The main test components were: helium to helium heat exchangers, hot gas ducts, hot gas valves, bellows, steam generators heliua , m purification system, helium blower theid an s r acoustic emissions, helium leakage, control othed an , r miscellaneous system componentsd an s e Th . overall operatio amountK KV nmore o stimt th f eeo than 20,00ever o nC ° 0highe hour0 95 t rsa temperatures. Its main operating parameters were:

~~Thermal power: 10 MJ/s (max. 12.8 MJ/s)~~

~~(maxC ° 0 . r 95 100105o t fo 0C 0C ° ° Temperature: short time periods)~~

~~Pressure: 40 bar (max. 46 bar)~~

~~Helium mass flow: max kg/4 . primarn so y helium-sid maxd ean . 20 kg/s on secondary helium-side s m/ 0 6 < Velocities:~~

~~Max. temperatur K/mi0 e20 transientn± s for test purposes:~~

~~Max. pressure transients + 5 bar/s for test purposes:~~

A test facilit Researce th t ya h Center Jiilich know EVA-Je th provides s na Jwa tess da t stanr dfo testing different types of steam reformer tube bundles and different catalysers for the steam reformin methanef go EVA-Ie Th . I test stand operate abour dfo t 15,00 (anr C ° 0fo d 0 hour95 t sa a smaller numbe hourf o r t 100sa 0 °C) mais .It n operating parameters were:

~~Thermal power: 10 MJ/s Temperature: 950 °C (max. 1000 °C) Pressure: 40 bar Helium maskg/se4 s flowc :~~

~~237 The main technical experiences relevant to the high temperature helium technology gained in the AYR and at the two related test facilities are summarized below.~~

~~1. Helium Leak Rate (AVR, KVK, EVA-II)~~

~~e followinTh g data were obtaine t normada l operation:~~

~~Nm5 inventorn 1 a 3t o t / d(a 0 1 f 200yo 0 Nm3)AVR: KVK: < 6 Nm3/d (at an inventory of 3600 Nm3) EVA-II: 2 - 3 Nm3/d (at an inventory of 1000 Nm3)~~

~~(KVKC ° 0 Ducts )95 Ga t t sa Ho 2.~~

Withi project P framewore nth PN basie e th ,th cs duct f developmeno ga k t s ho e th f o t including related construction elements, e. g., flange joints, bellows, bends, and different types of insulation, was performed in special smaller test stands. The integral testing at 950 °C to 1000°C was mainly performed in the KVK. The following test components for hot gas ducts were tested at steady-state and cycling transient conditions:

~~Test Time~~

Primary helium ducs systega tt mho 10100h ducs Primaga tt compensatoho r r 5300h connectios ga t core Ho th et n a outle t header 1700h Secondary helium system, hot gas duct 10600 h Ben r secondardfo y helium ducsystems ga tt ho , 10100h Axial-typ secondarn i 0 e 75 valv yW eN heliu m system 3900h Ball-typ secondarn i 0 e 70 valv yW eN heliu m system 5300h Raco-glob 0 e 15 valv W eN 5300h Axial-type valve NW 200 14000 h Hothrottls ga t 0 e 20 valv W eN 8700h

~~Main design feature experimentad an s l results:~~

reference Th e primar ducdesignes s ga wa tt yho coaxias da ducs ga t lheliu t ho wit e mhth in the inner tube and the cold helium in the outer ring gap. As flow guidance a liner consisting of a graphite tube and alternatively of a carbon-fibre-reinforced graphite tube were used insulatioe Th . n aroun outside linede th th rf e o consiste wrappea f do d fibre-type insulation (A1 SiO2d O3an 2 fibres graphitd an ) e foils arrange preveno dt t convective flows within the insulation. The pressure was retained by an outer, uncooled (no forced cooling) pressure tube.

The reference, secondary hot gas duct (900 °C) was designed as a duct with an inner flow guidance liner, an fibre-type insulation wrapped on the liner (A12O3 and SiO2) and an outer uncooled pressure tube.

~~Bends and bellows required for the primary and secondary hot gas ducts were also provided with similar insulation.~~

~~primar e secondare th th ductin s r r welga s fo yFo a t s la ygho protective coatings were used. They were provided as a sprayed-on base layer (Ni Cr Al Y), followed by a second~~

238 sprayed-on ZrO2-layer (stabilized wit 2 OhY 3 ). These coatings wer t positionea s where sliding motiono thermat e du s l expansions were expected e e courstesth th t f n o I e. operations (mounting and demounting of test sections), there were some spots were coating beed sha n damaged mechanically (locally chipped-off layer spots) thit sBu . damage y difficultie an t leadi no o d t d o seizin (n s r hookino g r similar)o g . Otherwise th e postexamination showed no damage or wear at the sliding supports.

mose Th t complicated component designee b o st bee d isolatioe dha n th ne valveth d an s regulation valvesecondare th n i s y helium ductdiffereno Tw . t type f isolatioo s n valves, namely axial-type and ball-type valves, were tested. The insulation was designed according to the same principle as for the straight secondary hot gas duct.

After experiening some initial difficulties, deficiencies wer t spotse ho repaire , ,g. . d(e defective supports, insufficient prevention of convective flows and defective temperature sensor penetratione th t a s s throug insulation)e hth performance Th . primare th f d o e yan secondar ductss ga t , yho includin g bend bellowsd san excellens wa , t even afte cycline th r g transient tests (e. g., 1500 load cycles for the bellows). Except the large hot gas valves, e desigprimare th th f secondard no s ductan y ga t s yho wit h flange joints, bellowd an s bends were considered as proven and reliable for a long-term operation at 950 °C (max. 1000 °C) desigA . yearn0 4 lif f seo migh expectede b t assuro T . e suc hlifa e time further endurance tests for the components are required.

appeart I s possibllesa e s expensivus o et e versio ductsf no , wit innen ha r metallic line. r(e g. made of Inconel 617) and no coaxial flow guidance for helium temperatures up to 950 °C or even 1000 °C. Moreover it became evident that a coating at the sliding transitions r slidino g supports (thermal expansions) migh t necessarilno t e neededb y finaA . l confirmation test is still required however.

The first tests of the large hot gas valves showed very unsatisfactory results concerning insulation, function and seat tightness. Although a large number of modifications leading essentiao t l improvements were made furthea , r developmen constructive th f to e desigd nan subsequent tests would be required before an application is possible.

~~3. Helium Purification and achieved Helium Atmosphere (AYR; KVK, EVA-II)~~

~~a) AYR~~

~~The AVR is provided with a helium purification system for removing radioactive and non-radioactive impurities t consistI . f threso e sections:~~

~~• A pre-purification section with a dust filter for particles > 0,3 /zm (throughput 1000 Nm3/h) followed by cooler (throughput 50 Nm3/h)~~

~~sectioA • r removanfo radioactivf o l e impurities (throughpu Nm0 5 t 3d /han )~~

~~• A section for removal of other non-radioactive impurities (throughput 50 Nm3/h)~~

purificatioe Th n syste mdesignes i thay tota e sucn di th twa lh a impurit y content (for all substances) at the purification outlet is < 10 ppm (for H2 < 1 ppm). The dust filters have a separation efficiency of at least 99,95 % for dust particles > 0,3 jim.

239 Practical experienc s showneha , that these specified requirements have been always fulfilled. No noteworthy difficulties have been experienced with the purification plant. Howeve surprisins wa t i r discoveo gt r that knowe onlth yf o aboun dus% 1 tt particles being distribute l possiblal n di heliuR e locationAY m e circuith f so t coul trappee db d in the pre-purification dust filter.

~~b)KVK~~

The KVK is provided with a redundant helium purification system. In additon to its classical layout, it is additionally equipped with an injection/doping system for injecting such substances as CO, CH4 and H2O into the helium circuit in order to adjust a defined helium atmosphere. Thi vers si y importan heliur fo t m systems operating C abov° 0 e85 because the metallic materials applied in high temperature helium systems are very sensitiv chemicae th o t e l attac heliuy kb m impurities (i.e., inner oxidatio r carno - burization, decarburizatio r formationo f stablno e oxide surfaces)e layerth n o e s Th . material tests have indicated that only a very defined narrow band of a permissible helium atmosphere composition can be tolerated and the permissible narrow band becomes smaller with increasing temperatures. In order to test the helium components in the KVK under realistic conditions, it was necessary to operate with a controlled helium atmosphere, i.e., with controlled impurities in the helium.

After having settled some problem commissioninge th t sa purificatioK KV e th , n plant generally operated very satisfactorily in spite of the difficult operating conditions of frequent opening resultine e systeth th d f mo san g ingres f airo s , humidity, dusd an t impurities vpm5 .othe< l ( )Excep al 2 H r r impuritietfo , H2 2N O s, 2 )(CH coulO , 4d,Co be purifie valueo dvppt 1 m< s each.

~~c) EVA-II~~

helium-purificatioe Th n plan EVA-Ir tfo layous Iha t simila tha o KVKr rt tfo , except that no injection/doping devic bees eha n provided operationae Th . l experiences correspond with KVK-plant e thosth f eo .

~~4. He-Blowers (AVR, KVK) a) AYR~~

The AVR is provided with two speed-controlled radial blowers (400 to 4400 rpm), each with a throughput of 13 kg/s. The blowers have operated at about 275 °C for roughly 130,000 hours without any noteworthy difficulties. Only in the case of a large water ingress in 1979, where water reached the blower axis, did the bearings of one blower became defective. The blower was withdrawn, impected, repaired and reinstalled.

~~Any sign f corrosioo s r erosio blowene o th n no r blades observed e coulb t dno .~~

~~b)KVK~~

~~The blowers provided in the KVK loops were required for achieving the helium circulation.~~

~~240 Table 3.3.4/1: Main Operationa BlowerK l DatKV f sao~~

Blower 1 2 Mass flow 3 3 kg/s Suction pressure 39 39 bar Suction temperature 230 230 °C Pressure increase 3.30 4.25 bar Rated speed 8660 15500 min "' Control range 20-100 20-100 % Design temperature 350 350 °C Design pressure 46 46 bar

The blowers are designed as compact uncooled radial compressors of horizontal barrel- type construction The installee yar steea n do l foundation wit integraten ha l systemdoi . Drive speed-controlled electric motors with intermediate gearing are provided.

~~Seals provided at the shaft penetrations of the compressor housing are labyrinth seals, floating ring gaskets, and a static seal.~~

The shaft wit e attachehth d impeller s supportei s lubricatey db d bearings arranged outside the compressor housing. The bearing housings are accessible from the outside without dismantling the main housing.

~~After having settled some initial problems (damage at the floating ring gasket and unsufficient function of the labyrinth seals) the blowers functioned very satisfactorily for more than 20,000 h.~~

As expected, at the rather low helium temperatures prevailing hi the blowers, no corrosion attack or deposits could be observed at the blower blades. Moreover no erosion attack coul visualle db y observed, althoug tese hth t condition beed sha n very rougd han dusothed an t r impuritie ingressed sha t numerouda s opening heliue th f smo systems.

~~The acoustic emmission of the two compressors into the environment had been very high at the commissioning. Improvements such as a sound insulation had to be made.~~

~~c) EVA-II~~

~~Similar experience describehavs K sa e KV beer dfo n made wit EVA-Ie hth I blowers.~~

~~5. Materials and Effect of Helium-Impurities (AYR, PNP, HHT)~~

The comprehensive German development and testing of metallic high temperature-resistant material applicatioe th r sfo heliun i m system heliud san m turbine desigt sa n temperatures up to 950 °C (max. 1050 °C) have shown the need for a careful control of the helium atmosphere at a narrow limited band for the permissible impurities.

~~241 At very high helium temperatures the following problems might arise:~~

~~• Inner oxidatio materiae th f no l leadin shortea o gt r lifetime~~

~~• Decarburization or carburization of the material leading to a shorter lifetime~~

~~• Destructio r formationo stabla f no e oxid esurface layeth n o re~~

~~Therefore the helium atmosphere with the impurities must be adjusted in such a way that stable protecting oxide layers form on the surfaces (without H-production) in order to~~

~~avoid an inner oxidation as well as a carburization or a decarburization2 , or other corrosion attacks.~~

The adjustment of the helium atmosphere occurs by purification and/or injection of defined substances, especially defined ratios of CO/H2O and CH4/H2O. It coul demonstratee db materiae th n di l tes t large standth en wels i a s tes s a lt facilities (KVK, EVA-II; EVO, HHV) that the permissible helium atmosphere can be orderly controlle adjusted dan heliue meany b dth f mo s purification system t vera d y an shig h helium temperature addtionae th y sb l provisio f injectiono n systems.

bloweR AY I nre bladescasth f eo corrosio o n , n erosio o attacn d knan were detectey db the visual inspection of one blower in 1979 (after the water ingress and the defective bearing event) considerin backgroune gth f knowdo largd nan e graphite dust quantities, AVR-helium-circuite th estimate n i e approx, b kg o dt 0 6 .observes wa t I . d that large quantitie graphitf so e dusbeed tha n stirre evert a p dyu startu loat a d pchangean other so r transients. A closer examination showed, that the dust consisted of graphite with only a few metallic particles or other substances. The largest portion of the dust was in the size range of 0.5 pm. However, some graphite particles were found with a size of 1500 /zm x 300 fjLm x 100 /itm and some metallic particles with a size of 400 fj,m x 200 /im x 50 /an. The visual inspection nevertheles t show no erosio y blowee d nan th sha n no r blades.

~~6. General Operation Experience largd an s e wit R High-TemperaturhAY e Test Facilities with Helium Circuit~~

~~wels Ii t l know reactoexcellenn R a nd AY tharha e tth t performance record durin- 20 s git year operation (approximately 127,000 hours) including 45,000 hours operation at 950 °C.~~

The KVK-plant was sucessfully operated for more than 20,000 h with peak temperatures up to 1050 °C and the EVA-II for more than 15000 h at 950 °C. Both test plants demonstrated that large sized, high temperature helium circuit operatee b n sca d safeld yan with a high availability (e. g. in the last three years of the test operation the KVK showed an availability > 95 %!).

~~242 Annex III~~

~~REFERENCES~~

~~[ l ] Bammert K. und Twardziok W.: "Kernkraftwerke mit Heliumturbinen für große Leistungen" Atomkernenergi (19672 1 e . 9/1No )0~~

~~[ 2 ] Bammert K., Krey G. und Küper D.: "Zusammenwirke Hochtemperaturreakton nvo Heliumturbined un r " Kerntechnik 11 (1969) No. 2~~

~~] 3 Bammer[ Rehbacd un . : K th J. "Gas turbine for a nuclear power plant" Atomkernenergie 18 (1971), No. 2~~

~~[ 4 ] Bammert K., Krey G. und Krapp R.: "Di MW-Heliumturbin0 e5 e Oberhause nAufba- Regelungd uun " Schweizerische Bauzeitun (19742 g9 )~~

~~[ 5 ] Bammert K. und Deuster G.: "Das Heliumturbinen-Kraftwerk Oberhausen - Auslegung und Aufbau" Energi Technik+ e , Issu , 197l e 4~~

~~] [6 ZenkerP.: "Die Heliumturbin r Kraftwerkstechnikde n ei " Wärme, Band 78, Issue 1/2, 1972~~

~~[7] ZenkerP.: "Die 50-MW-Heliumturbinenanlage Oberhausen" VGB-Kraftwerkstechnik, 55th-Annual, Issu , 19711 e 5~~

~~] 8 Zenke[ : P. r "Einige Aspekte zum Bau der 50 MW-Heliumturbinenanlage der Energieversorgung Oberhausen" Atomkernenergie 23 (1974), No. 2~~

~~] [9 ZenkerP.: "Das Helium-Kraftwerk Oberhause Kraftwern ei n- r Zukunftde k " r StädtetaDe (1975g6 )~~

~~] Deuste0 Plüd : 1 [ un R. r . G r "Betriebserfahrunge t Heißluftkraftwerkemi n d Folgerungeun ne weiterdi r fü en Entwicklung von Heliumkraftwerken" Mitteilunge(1971)1 5 2 . B No , VG r nde~~

~~[11] Plür R. und Wilzhoff H.: "Di MW-He-Gasturbinenanlag0 5 e e Oberhausen" Chemie-Technik (1976)0 ,6 Pag- 5 e5~~

~~243 l 2 GHH1 [ , Oberhausen "Heliumturbine r Kernkraftwerkenfü " GHH - Technical Reports 4/67~~

~~] 3 GHH1 [ , Oberhausen "Gasturbine geschlossenem ni n Prozeß" GHH-Brochure M.TM 809d677os~~

~~] GHH4 1 [ , Oberhausen "Helium-Gasturbinen-Anlag Oberhausenn ei " GHH-Brochure So432 (M.TM) 8dl075~~

~~[ 15 ] Griepentrog H.: Heliumturbinen-Technologie" GHH-Brochure 432So (M.TM) 13d76br~~

~~] Stor6 1 [: zR. "Erfahrunge r EVO-Helium-Turbinen-Anlagede t nmi " Technischer Bericht 77.102, Interatom August 1977~~

~~[ 17 ] Noack G., Weiskopf H.: "Die Hochtemperatur-Helium-Versuchsanlage (HHV) - Aufbau und Beschreibung der Anlage" Jül 1403, March 1977~~

~~[ 18 ] BBC "Die Hochtemperatur-Helium-Versuchsanlage (HHV-Anlage)" BBC-Brochure D KW 507 36 D~~

~~[19] ZenkerP.: " 10 Jahre Betriebserfahrung mit der Heliumturbinenanlage Oberhausen" VGB Kraftwerkstechnik, Issue 7/1988~~

~~[20] ZenkerP.: "Ausgewählte Rohrleitungsprobleme der Heliumturbinenanlage Oberhausen" "3R international", Issu , 197e4 6~~

~~[21] Trable: M. r "Dokumentation Heliumturbinenanlage Oberhausen" KWU, Technical Report VK 91/87/16, Erlangen, 1987~~

] 2 "Untersuchun2 [ r Wellenschwingungede g s Niederdruckverdichterde n r de d un s Hochdruckgruppe an der Heliumturbinenanlage der Energieversorgung Oberhausen" Abschlußberich Versuchsvorhabens de t , 198A s 1 Institut für Mechanik der Universität Hannover

~~[ 23 ] HHT-Projekt "Forschungs Entwicklungsarbeitend un - " Abschlußbericht übeZeitraun rde m 1978 - 198 1 April 1982~~

~~244 [ 24 ] HHT-Projekt "Ergebnisse der Entwicklung und Planung des Hochtemperaturreaktors mit Heliumturbine von 1969 bis 1982" March 1983~~

~~] Bammer5 2 [ , PösentrupK. t : "Das Verhalten einer geschlossenen Gasturbin i zeitlicbe e h veränderlichen Betriebszuständen" VDI-Forschungsheft 588/1978~~

] Bammer6 2 [ , JohanninK. t , WeidnegL : E. r "Instrumentierun d Verfahreun g r Messunzu n s dynamischede g n Verhaltenr de s Heliumturbinenanlage Oberhausen" Atomkernenergie, Kerntechnik, Vol , 19834 .. 43 No ,

[ 27 ] Bammert K.: "Operating Experience Measurementd san Turbn so o Set CCGT-Cogeneratiof so n Plants in Germany" ASME-Pape . 86-GT-10No r 1 31st International Gas-Turbine-Conference 1986, Düsseldorf

~~[ 28 ] Personal Communication with Poggemann R. und Zahn G. EVO, Oberhausen, September 1994~~

~~[ 29 ] Personal Communication with A. Bald, Siemens/KWU hi September 1994~~

~~[30] ThodeK. H.: "Schall- und Schwingungsmessungen an der Heliumturbinenanlage bei der Energieversorgung Oberhausen" Interatom, Teil des Schlußberichtes des Versuchsvorhabens B, 1982~~

[31] ÖryH.: "Dehnungsmessungen am "Innenliner" der Heliumturbinenanlage der Energieversorgung Oberhausen" TeiSchlußberichtes de l Versuchsvorhabens sde , 1981B s , Institu Leichtbar fü t r Technude . Hochschule Aachen

~~[32] "Nickel: W. . "Untersuchung der Druck- und Temperaturbelastung der Innenisolierung der Koaxialleitung an der 50 MW Heliumturbinenanlage in Oberhausen" Interatom, 1979~~

[ 33 ] Ebers A.: "Untersuchung des Isolationsverhaltens der Innenisolierung der Koaxialleitung an der 50 MW Heliumturbinenanlage bei der Energieversorgung Oberhausen" Interatom, Schlußberich Versuchsvorhabens de t , 198C s 1

~~[ 34 ] Hebel G., Weiskopf H., Terkessidis J., Hofmann W.: "HHV-Anlage; Erfahrungen bei Bau und Inbetriebnahme" Atomwirtschaft/Atomtechnik Issue No.5, May 1982~~

~~245 [ 35 ] Personal Communication with Arndt E., Weiskopf H., König H. H., ABB Mannheim in Septembe Octobed an r r 1994~~

] Persona6 3 [ l Communication with Kugele , BarnerK. r , FröhinH. t , gW. Terkessidi , BröckerhofJ. s , GottauP. f , SingH. t , SchuberhJ. , F. t Schuster H. et al. KFA Jülich in August/September/October 1994

~~[ 37 ] ABB Review Issue 8/9; 1989~~

~~B [38AB ] "Zeitgemäße, umweltfreundliche Kraftwerkstechni t Gasturbinenkmi " ABB-Brochur 207, MarcT D PG 2 1 9 . h 199No e 2~~

~~[39] Schulten R.: "Die derzeitige Situation der Kernenergie und ihre zukünftigen Aussichten" Atomwirtschaft, February 1988~~

~~] AVR-He-Reinigungsanlag0 4 [ e (Auszu Sicherheitsberichts gau ) AVR~~

~~[41] Wahl!.: "Die Kontaminatio AVR-Primärsystems nde s nach Jahre 0 meh1 s ral n Reaktorbetrieb" AVR-Report, March 1980~~

] Hannlik2 4 [ , Iven , KrügesG. , ReimeK. r , ScheubeL r , SchmieE. r , WahdH. : J. l "Plate-Out-Messungen und Dekontaminationsarbeiten an Bauteilen des AVR in Jülich" EIR-Report No. 384, January 1980

~~[ 43 ] AVR/BBC/HRB "Der Kugelhaufenreaktor der Arbeitsgemeinschaft Versuchsreaktor (AVR)" Brochure D KW 1161 87 D, May 1987~~

~~] BBC/HR4 4 [ B "Langzeiterfahrunge AVR-Versuchskraftwerkm de t nmi " BBC/HRB-Brochur 116D W K 7 78 D e~~

~~[ 45 ] Wawrzik H., Biedermann P., Oetjen H. F.: "Staub im AVR-Reaktor; Verhalten bei transienten Strömungsbedingungen" Vortrag Jahrestagung Kerntechnische Gesellschaft 1988~~

~~R AV ] 6 4 [ "AVR - 20 Jahre Betrieb" VDI-Repor . 729 198y No t ,Ma 9~~

~~] Persona7 4 [ l Communication with Wimmers,R PohAV - l in September/October 1994~~

246 [ 48 ] "Metallische Werkstoffe für die wärmetauschenden Komponenten und hochbelasteten Einbauten von Hochtemperaturreaktoren zur Erzeugung nuklearer Prozeßwärme" Statusseminar January 1983 at Ministerium für Wirtschaft, Mittelstand und Verkehr, North-Rhine-Westfalia

] HTR-Komponente9 4 [ n "Stand der Entwicklungen auf dem Gebiet der wärmeführenden und wärmetauschenden Komponenten" Symposium January 1984 at Minsterium für Wirtschaft, Mittelstand und Verkehr, North-Rhine-Westfalia Volume I and II

~~[ 50 ] Hochtemperatur-Kreislauf - Interatom/Siemens-Brochure October 1983~~

~~[ 51 ] IAEA-Conference "Specialist Meeting on Heat Exchanging Components of Gas-Cooled Reactors" April 198 t Bon4a n (BMFT)~~

[52] Jansing W.: "Testing of High Temperature Components in the Component Testing Facility (KVK)" Vortrag VGB-Sondertagung "Kohleumwandlun Hochtemperaturreaktord gun " Interatom/Siemens, September 1985

~~[ 53 ] Fröhling W., Ballensiefen G.: "The High-Temperature Reactor and Nuclear Process Heat Applications" Nuclear Engineerin Designd gan , Volum ( 1984)2 8 . e7 No ,~~

~~[ 54 ] Nukleare Fernenergie Zusammenfassender Berich Projekm zu t t Nukleare Fernenergie (NFE) KFA Julien/Rheinische Braunkohlenwerk KölG eA n Jül-Spez-303, March 1985~~

~~] Sicherhei5 5 [ Hochtemperaturreaktoren vo t n KFA/KTG Jül-Conf , Jun53 . e 1985~~

~~[ 56 ] Harth R., Jansing W., Mauersberger R., Teubner H.: "Stan r de Komponentenentwicklund n Hochtemperaturreaktode r fü g r zu r Prozeßwärmeerzeugung " Atomkernenergie-Kerntechnik Issue 3/1985~~

~~[ 57 ] Jansing W., Breitling H., Candeli R., Teubner H. (Interatom): "KVK and Status of the High Temperature Component Development" IAEA-Conferenc Jülichn i e , October 1986~~

~~[58] Hundhausen W.: Endbericht zu den Untersuchungen zum Isolationsverhalten der Sekundärheißgasleitung im KVK Interatom/Siemens, August 1987~~

247 ] Brandstette9 5 [ , JansinA. r: gW. Ergebnisse des PNP-Projektes - Vertrag VGB-Sondertagung "9. Internationale Konferenz über den Hochtemperaturreaktor - Kohle und Kernenergie für die Strom- und Gaserzeugung" Interatom/Siemens, October 1987

~~[ 60 ] Fräsen: Abschlußbericht Test Primärkompensator Interatom/Siemens, November 1988~~

~~[ 61 ] StausebachD.: " Entwicklun Heißgasleitungen gvo n für Hochtemperaturreaktore Temperaturbereicm ni h " °C 0 95 s bi Siemens/Interatom, June 1989~~

~~[62] Hundhausen W.: Endbericht zu den Untersuchungen am Zwischenkreislauf-Rohrbogen im KVK Interatom/Siemens, November 1989~~

~~[63] Jansin , TeubnegW. : H. r "Der Hochtemperatur-Heliumkreislauf KVK - Erfahrungen aus 20.000 h Versuchsbetrieb" Jahrestagung Kerntechnische Gesellschaft 1990 Interatom/Siemens~~

~~] Hart4 6 , Jansin[ hR. , TeubnegW. : H. r "Experience Gained OperationfroK EVA-Ie mth KV d an I " Nuclear Engineerin Desigd gan (19901 n12 )~~

~~[65] Schubert F., Rittenhouse P.: "Metals Technolog r Modulayfo r HTG Turbiness Ga R- " International Workshop, Boston, June 1991~~

] Nicke6 6 , [ HofmanH. l , WachholnK. , WeisbrodzW. : I. t "The Helium-cooled High Temperature Reacto FRGe th i h r: Safety Features integrity Concept, Outcoo Desigr fo l n Code Licensind san g Procedures" Nuclear Engineerin Desigd gan (19917 n12 )

~~[ 67 ] Personal Communication with Jansing W., Teubner H. et al - Siemens (Interatom) in August/September/October 1994~~

~~248 WORKSHOP~~

~~Session 5 PEBBL MODULAD EBE R REACTOR - SOUT H AFRICA~~

~~M. FOX Integrators of System Technology, "l""1" XA9642790 Waterkloof~~

~~E. MULDER ESKOM~~

~~South Africa~~

~~Abstract~~

In 1995 the South African Electricity Utility, ESKOM, was convinced of the economical advantages of high temperature gas-cooled reactors as viable supply side option. Subsequently planning of a techno/economic study for the year 1996 was initiated.

~~Continuation constructionthe to phase prototypea of plant will depend entirelyoutcomethe on thisof study.~~

A reactor plant of pebble bed design coupled with a direct helium cycle is perceived. The electrical output is limited to about 100 MW for reasons of safety, economics and flexibility. Design of the reactor willbasedbe internationallyon proven, available technology. extendedAn researchand development program is not anticipated.

~~New licensing rules and regulations will be required. Safety classification of components will be based on the merit of HTGR technology rather than attempting to adhere to traditional LWR rules.~~

A medium term time schedule design the construction for and prototypea of plant, commissioningand performance testing is proposed during the years 2002 and 2003. Pending the performance outcome this of current the plant and power demand, series units production foreseen.is MWe 100 of

~~UTILITY INTEREST~~

Load demand forecasting in South Africa includes a fossil-fired power plant under construction as well actionin e de-mothballingth numbea f go f out-of-servico r e boilers worsa n I . t case load growth scenario (3-4% loa dreserv % growth20 e e th )margi n wiljeopardizee b l yeae th ry db 2006 . This suggests that serious consideration mus immediate givee th b t o nt e plannin f additionago l power plant. Should ESKOM, South Africa's single largest utility, decide to construct a new large-scale coal plant, they would commit to a substantial capital intensive project with a construction lead time of 5-8 years. Economic viability dictates that suc planha locatee b t closdn i e proximit coae th l o yt field n so the arid Highveld area latese Th . t plant under constructio fittes ni d wit cooliny hdr g towers, hence adding to the enormous capital investment. abovementionee Keepinth f o l gal mindn di logicaa , l solution woulbuilo t e ddb lower capital cost intensive power plants, with 18 to 24 months' construction time and the freedom to be erected at coastal regions or as single plants in more remote areas eliminating the considerable costs associated with transmission. If the unit capital cost of such plants compare favorably with the bigger plants, then this would certainly become part of ESKOM's supply side options.

Indications are that modular, high temperature, helium-cooled, direct cycle nuclear power plants could be viewed as a strong contestant, hence the willingness to support a feasibility study to establish the PBMR viability.

251 South Africa ESKOd an , Mparticulan i r recently completed studie alternativn so e electricity supply side options e.g. gas, hydro, win solard dan . Large, constant flowing river lackine sar r producingfo g hydro-electric power thas tGa . mus importee b t d from neighboring countries prove highle b o dt y uneconomical. Wind, even at the coastal areas is insufficient for economical power production on large scale. The South African government, however, recently decided to spend a considerable amount of money on the supply of solar units to schools, clinics, and individual households, etc., in remote lying regions, such as Kwazulu-Natal.

Political perceptions politicalle Th y sensitive natur f nucleaeo r projects fro mpublia c acceptance poin f vieo t welws i l acknowledged. Open publicity right fro projece onsee mth th f o tt shoul handlee db publia y db c relations team, supporte technicae th y db l expert co-operation i s n witcliente hth . Thi s largelsi y possible due to the local need for electricity supply, the safety characteristics of the high temperature helium-cooled reactor and the economic advantages offered by the direct cycle power conversion. Support is found in a list of potential benefits, such as high local content, export development, desalination, exploitation of the existing skills base, minimal environmental impact, etc. The initial goal is to employ this beautiful technology to benefit the South African population at large. Mounting public awareness of the fact that the environmental impact of power sources derived from coal energy, is not negligible, contributes towards greater acceptance of nuclear as energy source.

~~Support wilsoughe b l t from South Africa's Departmen f Mineralo t d Energan s y e Affairth r fo s proposed techno-economic study.~~

~~TECHNO-ECONOMIC Study 1996~~

The decisio basno t cosea t evaluatio preliminara n no y design rather than derivin t gfroi m costing models was derived from the fact that cost references are firstly, not direct cycle specific and secondly, deviates from existing design bases.

1ST will compile a works proposal document identifying work packages to cover a broad spectrum of disciplines from the fuel acquisition right through to the civil structures. Indications of the limited funding available, will limit the depth of investigation. It is sincerely believed that expert involvement from all over the world in basic design, reviews, comments and costing, will however, enable arrival at a sound technical proposal with realistic cost estimates.

~~OWNE USED RAN R REQUIREMENT SPECIFICATION~~

~~The stud strongls yi y drive technicay nb financiad an l l objectives. ESKOM recently "publishede th " first draft of two PBMR-specific reports, namely:~~

~~• User's Requirement Specification (URS), and OwnersA • ' Requirement Specification (ORS).~~

~~These documents specify r examplefo , , reactor availability, reacto targed an r t unit capitad an l operational costs.~~

The mai onld nyan real objectiv r thiefo s year's techno-economic studmeasuro t s i y resulte eth s against the objectives as spelt out in the ORS/URS. Certainly a first prototype plant will cost more tha unie serienn th i t s production, therefor costine eth g design e wildone th b lr efo , prototype plant serier costfo d s san productio n plants same .Th e applie fuelr sfo .

~~252 REACTOE !00MW R CONCEPT~~

~~The anticipated plant main characteristics are:~~

~~• Performance~~

Cycle Brayto= n direct cycle (recuperated) Power = 200 -> 220 MWth; 100 MWe Overall efficiency Bette= r tha% n45 Fuel = Triso particles in 60 mm diameter spheres Enrichment < 20% Power density MW/m4 » - 3 3 = Reactor Temperature = 900 °C outlet, 604 °C inlet

Average burnup = 130MWd/kgHM Maximum allowable fuel temperature = 1600 °C Maximu mase mH s flow rate Systea MP kg/ 0 m7 t 13 sPressura = e Pressure ratio = 2.2 (2 stage compression with intercooler) Fuelling Continuou= s Defuelling Batc= h (once ever years3 y2- ) Generator speed 300= m 0rp Turbo/compressor speed = 12000 rpm Decay heat removal = Passive Control System = Non safety class. Industrial high standard with redundancy.

~~Physical~~

Core height determinee b o T = d Initial height = 5,8 meters Maximum height = 9 meters Core diameter = 3,6 meters Number of control rods = 12 Control rod system = Absorber rod + absorber chain Suppor f rotationao t l parts = Magnetic Building size = L = 40m, W = 20m, H = 40m Position = Choice of depth based on economics /safety.

~~Safety~~

~~activo N e safety system~~

~~Seismic design as required~~

Protection against external events (e.g. in case the reactor pit area above ground level, the confinement should be bunkered) No off-site emergency plan (preliminary investigation indicatese thib so t possible).

~~Plant Layout~~

Indication thae reactore sar th t , PCD, secondary systems, etc. ,e auxiliar mosth f o t y controd an l instrumentation systems could be housed within a concrete structure of size 40mx20mx40m. The primary components could be housed in a protective reactor pit. One of the design packages should include the finalization of design criteria for these components and housing.

~~253 A manifold would be coupled to the reactor. The power conversion components are then coupled to this manifol primaro t r do y pipework located insid manifolde eth .~~

~~The Reactor core and Fuel~~

The proposed reactor design woul HTR-Modubasee e db th n do l reactor t coulI . d simpl describee yb d a pebblsa (sphered ebe ) core with graphite reflector, carbon insulation, steel core steebarrea d l an l pressure vessel maie Th .n deviations fro Modue mth l reacto antcipatee rar : be do t higheA • r outlet temperature. • Design provision must be made for a higher gas return temperature. • Flow channel proposee sar d insid carboe eth n insulation. • Contro drivd ro l e mechanisms mounted externall reactoe th o yt r pressure vessel. • A separate shut down system could be excluded. • Provision woul made db off-linr efo e batch defuellin on-lind gan e peti-a-peu fuelling. • Reactor vessel coul coolee d b col a y ddb helium leak strea sideU m .froPC e mth • Core provided with nibs to enhance heat decay heat transfer to passive cooling system. • Average core power density and average burn up anticipated to be higher.

Reactor isolation, reactor support, core internal packaging method, etc., will be similar to the Modul reactor Germae fuee th f Th l.o wilne b ldesign source Th . e wily available b l ma worle d th dan n ei eventually be partly manufactured in South Africa.

~~Power Conversion Unit (PCU Generatod an ) r~~

The PCU main components, including the generator is anticipated to operate inside a 70 bar pressure boundary. This pressure boundary consist couplina f so g vessereactoe th o t l r (calle manifolde dth ) separatano dtw e pressure vessel tower bells)4 ( s turbo-compressor e Th . intercooled san r wile b l house towee th n dri neares reactoe th o t t r whil powee eth r turbine, recuperato pre-cooled an r e ar r housed in the other tower.

~~Interesting characteristics (features) in the proposal are as follows:~~

• Separate turbo machinery shafts. • Power turbine operate pressurew lo n temperaturso w lo , e side. • Differential pressure acros primare sth y pipin lowgs i , especiall hottese yth t pipes. • Small leakage on piping allowed for. • The pressure boundary is low temperature (can be left uninsulated or use cold He leak as coolant). • Thermal expansion unconstrained. • Differential thermal expansion addressee.ge pipeb n o n . sca d with bellows. • Watesystee th n ri m alway lowet sa r than system pressure during operation. • Access to the first turbo compressor and generator relatively easy. Other components also accessible. • Isolation between the reactor and PCU allows for enhanced access to PCU components. • individuae Sizth f eo l components makes manufacturin handlind gan g easier. • Testin qualificatiod gan individuaf no l component relativels i y simple. • Easy replacemen componentf o t s (especiall prototype th n yi e plant possibles i ) . • Startup via a jet pump system. This allows for startup of the system in remote areas where the electricity gri unavailableds i .

~~As is previously mentioned, all operating and accident scenarios may not be accounted for in the current engineering study.~~

~~254 Secondary systems~~

~~Of the many secondary systems, some brief comments are made of:~~

~~The passive heat removal system: believe W e tha specifia t c active residual heat removal system requirede b ma t yno .~~

Defuelling system: Shoul locatee db d externall reactoe th o yt r vessel defuelline Th . g chute must physicall shors a possible s e a t yb respective Th . e design proposal Germann si Chind yan a wile b l investigated.

Spent-fuel system: The use of an intermediate spent-fuel storage system consisting of multiple containers, coole y naturadb r circulatioai l s envisagedni . Remova containere th f o l s e neeb o dt investigated.

Primary loop clean-up system: Preliminary investigation in-linn a f so e suggese filterus e . Activth t e gasses will be adsorbed by dust particles and subsequent plate-out is anticipated on the cooled surfaces of the heat exchangers. A percentage of the gas will be filtered out in an anticipated bypass system presence Th . gaseoua f eo s waste storage system must stil investigatede b l .

~~Contro Instrumentatiod an l n~~

The control system is anticipated to be of typical high standard industrial type. Redundancy will be provided where required. Overall classificatio controe th f no l syste manticipates i non-safete b o dt y graded.

~~LICENSING AND REGULATION~~

An overall licensing and regulatory basis appears to be lacking. This is an issue which requires in- depth resolution in the medium term. It is anticipated that the utility will undertake to develop in-house the regulatory guidelines.

Documentation indicating the licensing process as well as providing a basis for licensing (modified 10 CFR 50/52) must be put in place. Discussions with the local Council for Nuclear Safety (CNS) is to be scheduled on a continuous basis.

~~International contact wil establishee b l d with relevant institution companiesd san .~~

~~TECHNOLOGY~~

Sout hr man Africfo s y aha year s been involve nuclean di r related projects. w Hig lo wels ha s a l enrichment plants were designed, built and operated. Fuel had been manufactured for the Koeberg Nuclear Powe SAFARe th r r fo Plan Id materialan t s test reactor. Isotopes, suc Molybdenum-9s ha 9 are currently being produce ongoind dan g researc developmend an h t wor lasen o k r enrichment technology forms part of the Atomic Energy Corporation's (AEC) present activities. A center of excellence on small turbines and small to large compressors resides within the AEC. Reactor studies have previously been conducted including smal mediuo t l m powe R sizePW r smalplanta d slan peu- a-peu HTGR.

~~255 TIME SCALE~~

~~«- Des;ign -» Construct Commercial Use Study <-Comp. Mar ufacturing-> <- Comp. Testing -> -> 1996 1997 1998 1999 2000 2001 2002 2003~~

~~256 TH IAEEE ROLGAS-COOLEN ATH I F EO D REACTOR XA9642791 DEVELOPMENT AND APPLICATION~~

~~J CLEVELAND BREYL , J ,KUPIT Z Divisio f Nucleano r Power, International Atomic Energy Agency, Vienna~~

~~Abstract~~

Within the Statute establishing the International Atomic Energy Agency there are several functions authorized for the Agency. One of these functions is "to encourage and assist researc developmend an , hon practicad tan l applicatio , atominof c energ peacefur yfo l uses throughout the world...". The development of nuclear power is deemed an important application of this function. The representatives of Member States with national gas cooled reactor (GCR) programmes advise the Agency on its activities in the development and applicatio GCRe committeth e f nTh o . technologf R leadereo GC n si y representing these Member States is the International Working Group on Gas Cooled Reactors (IWGGCR).

The activities carried out by the Agency under the frame of the IWGGCR include technical information exchange meetings and cooperative Coordinated Research Programmes. Within the technical information exchange meetings are Specialist Meetings to review progress on selected technology areas and Technical Committee Meetings and Workshops for more general participation. Consultancie Advisord an s y Group Meeting convenee ar s o dt provide the Agency with advise on specific technical matters. The Coordinated Research Programmes (CRPs) established within the frame of the IWGGCR for the GCR programme include:

* Validatio f Safetno y Related Physics Calculation Enrichew Lo r sfo d GCRs, * Validation of Predictive Methods for Fuel and Fission Product Behaviour in GCRs, * Heat Transpor Afterhead tan t Heat Remova GCRr lfo s under Accident Conditions, and * Desig Evaluatiod nan f Heano t Utilization Systeme Higth r h fo sTemperatur e Engineering Test Reactor.

~~This paper summarizes the role of the International Atomic Energy Agency in GCR technology development and application.~~

~~1. Introduction~~

Internationae Th l Atomic Energy Agency (IAEAfunctioe th s )"fosteha o n t exchange th r f eo scientific and technical information", and "encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world".

IAEe Th Aadvises i activities it n do developmenn si applicatiod an t gas-coolef no d reactors by the International Working Group on Gas-Cooled Reactors (IWGGCR) which is a committee of leaders in national programmes in this technology. The IWGGCR meets periodically to serve as a global foru informatior mfo n exchang progresd ean s report nationae th n so l programmes identifo t , y area collaboratior sfo adviso t d IAEe programmes neit an th n Ao . This regular revie conductews i d opein na n foru whicmn i h operating experienc developmend ean t programme frankle sar y discussed. Countries participating in the IWGGCR include Austria, China, France, Germany, Italy, Japan, the

257 Netherlands, Poland Russiae th , n Federation, Switzerland Unitee th , d Kingdo Unitee th d mdan States of America additionn I . OECD-NEe th , Europeae th d Aan n Union participat IWGGCRe th n ei .

~~This paper describe e IAE th e rol f n Gas-Cooleth Asi o e d Reactor (GCR) technology development and application.~~

~~. 2 Background~~

Worldwid largea e amoun experiencf to bees eha n accumulated during development, licensing, construction and operation of gas-cooled reactors. The experience forms a sound basis for programmes which are underway in several countries to develop advanced high temperature reactors for electric power generation and for process heat.

~~2.1. Summary of operating experience~~

In the United Kingdom approximately 937 reactor years of operating experience with carbon dioxide cooled reactors has been achieved*". Over 20% of the UK's total electricity is generated by its 20 Magnox and 14 AGR gas-cooled reactors, with the AGRs achieving a combined average annual load factor of 75.6% in 1994, the highest of all reactor types worldwide. This remarkable improvement relative to the earlier performance resulted from successful efforts by Nuclear Electric to reduce trip rate outagd san e times improvo t , refuelline eth g procedure increaso t d san e thermal efficiencies. However, no further OCRs are planned hi the UK, and development work will be concentrated on further improvements in plant performance and life extension of existing plants.

~~(1) based IAEAn o PRIS data base includingd an smalle th - 50MW(e)}{ Colder HallChapeld an Cross units.~~

In France, about 200 reactor years of experience have been acquired through operation of eight Magnox-type reactors demonstratin soundnesse gth , fro mtechnicaa safetd an l y poin viewf o t , of this reactor technology. However, the decision was made some time ago to concentrate on large pressurized water reactors, and the last of France's Magnox reactors, Bugey 1, was shutdown in 1994.

~~In Japan the 159 MW(e) Tokai-1 Magnox-type reactor continues to be a very successful plant.~~

The experience with the early helium cooled High Temperature Gas-cooled Reactors (HTGRs), the Dragon plant in the UK, the AYR in Germany and Peach Bottom in the USA was very satisfactory experience Th . e wit latee hth r HTGRs, For . Vraie St tth d n (33an A 0 MW(e)US e th n )i THTR-300 (300 MW(e) Germanyn i )t entirel no s ywa , satisfactory probleme Th . s which resulted shutdowe inth f thesno e plants were, however t relatebasie no ,th co dt reactor concep f heliuo t m cooling, and the use of graphite for neutron moderation and as a structural material, nor were they related to any safety concerns, but were primarily associated with technical and economic problems with first-of-a-kind system componentsd san .

~~2.2. Summary of national HTGR programmes~~

~~Active technology development programme HTGRr sfo proceedine sar Chinan gi , Japad nan the Russian Federation.~~

In Japan an important milestone in development of gas-cooled reactors was reached in March 1991 wit stare hconstructioth f to Hige th hf n o Temperatur e Engineering Test Reactor (HTTRe th t )a Oarai Research Establishment of the Japan Atomic Energy Research Institute (JAERI). This 30

258 MW(t) reactor will produce core outlet temperatures of 850 °C at rated operation and 950 °C at high temperature test operation firse th t wil I t .e nucleab l r reacto worle connectee th b n o di rt higa o dt h temperature process heat utilization system. Criticality is expected to be attained in 1998. The reactor will be utilized to establish basic technologies for advanced HTGRs, to demonstrate nuclear process heat application, and to serve as an irradiation test facility for research in high temperature technologies. The timely completion and successful operation of the HTTR and its heat utilization system wil majoe b l r milestone gas-coolen i s d reactor developmen developmenn i d an t f nucleao t r process heat applications.

In China Russiae th , developmennA FederatioUS e th d tnan effort electricitr fo s y producing systems concentrat smaln eo l modular HTGR designs with individua0 l powe28 o t r0 8 rating e th n si MW(e) range. Strong emphasis is placed on achieving a high level of safety through reliance on inherent features and passive systems. Satisfying this objective forms the basis for the smaller power output of individual modules and for the reactor core configuration. Emphasis has also been placed on a maximum use of factory fabrication, as opposed to field construction, for better quality control and reduction in construction time.

A promisin approacw gne achievho t e economic advantage modula e involveth f o e r sHTGus R with a gas turbine to achieve a highly efficient electric generating system. Recent advances in turbomachinery and heat exchanger technology have led to plant design and development activities in the USA and Russia, with the direct helium cycle as the ultimate goal. It is recognized that the unique feature modulae th f o s r HTGRs will likely require prototype demonstration prio desigo t r n certificatio commercializationd nan . Wit relativele hth y small siz eacf eo h power-producing module it is possible to contemplate such a demonstration with just one module, later expanding into a multi-module plant at the same site for commercial purposes. A Technical Committee Meeting on "Design and Development of OCRs with Closed Cycle Gas-Turbines" is scheduled for 30 October to 2 November 1995 at the Institute for Nuclear Technology, Tsinghua University in Beijing, China. A decision was recently made by the USA to focus on the ALWR concept and close out their GCR activities.

China' developmenR HT s t activitie MW(th0 1 e focusee ar sth ) n Teso d t Module HTR. Construction of the HTR-10 Test Module began in late 1994 at the Institute of Nuclear Energy Technology of Tsinghua University in Beijing. This project will provide experience in design, construction and operation of an HTR. The test module is designed for a wide range of possible applications, for example, electricity, steam and district heat generation in the first phase, and process heat generation in the second phase.

In Germany a strong HTR technology programme was performed in the 1970s and 1980s, desigR HT annn a dwit ha ver y high degre f safeto e s beeha yn developed bot r electricithfo y generation and for process heat applications. Inherent features and properties of HTRs are particularly conducive to achieving a nuclear technology that is "catastrophe free" and extensive research, developmen demonstratiod an t n activities have been conducte procesy ke n do s heat plant components. The helium heated steam reformer, the helium/helium heat exchanger and the helium heate generatos dga coar rfo l refining have been successfully teste pilon di t scale (e.g.d MW)an 0 1 , , the AYR reactor has demonstrated operation at 950°C core outlet helium temperature.

In Switzerland, in the past, research activities for small HTR concepts including the gas- cooled district heating reactors have been conducted. Current HTR-related activitie Switzerlann si d involve the PROTEUS critical experiments which are being conducted by an international team of researcher Paue th lt sa Scherre r Institut Villigenn ei . Activitie underwae sar Netherlande th n yi o t s assess the potential future role of modular HTRs as a highly safe technology for electric power generation. Other countries including Poland, Italy, Indonesia, and Israel have displayed interest in HTR technolog perford yan m related assessments.

~~259 3. International Cooperation~~

earle Th y developmen nucleaf o t r powe conductes rwa larga o dt e exten nationaa n o t l basis. However, for advanced reactors, international co-operation is playing a greater role, and the IAEA promotes international co-operation in advanced reactor development and application. Especially for designs incorporating innovative features, international co-operatio plan importann nca a y t role allowing a pooling of resources and expertise in areas of common interest to help to meet the high cost f developmentso .

To support the IAEA's function of encouraging development and application of atomic energy for peaceful uses throughout the world, the IAEA's nuclear power programme promotes technical information exchang co-operatiod an e n between Member States with major reactor development programmes, offers assistance to Member States with an interest in exploratory or research programmes published an , s reportcurrene th n so t statu reactof so r development whic available har e l Membetoal r States.

activitiee Th IAEe sth carrieAy b t withi dou IWGGCe fram e th nth f o e R include technical information exchange meeting co-operativd san e Co-ordinated Research Programmes (CRPs). Small Specialists Meetings are convened to review progress on selected technology areas in which there is a mutual interest. For more general participation, larger Technical Committee Meetings, Symposia or Workshops are held. Further, the IWGGCR sometimes advises the IAEA to establish international co-operative research programmes in areas of common interest. These co-operative efforts are carried out through Co-ordinated Research Programmes (CRPs), are typically 3 to 6 years in duration, and often involve experimental activities. Such CRPs allow a sharing of efforts on an international basis and benefit from the experience and expertise of researchers from the participating institutes.

IAEA'e Th s activitie gas-coolen i s d reactor developmen tfoue focuth r n technicao s l areas whic predictee har provido dt e advanced HTGRs wit higha h degre f safetyeo t whicbu , h muse b t proven. These technical areas are:

safe th e neutroa) n physics behaviou reactoe th f ro r core b) reliance on ceramic coated fuel particles to retain fission products even under extreme accident conditions c) the ability of the designs to dissipate decay heat by natural heat transport mechanisms, and safe th e behavioud) reactofued e th an l f ro r core under chemical attack (ai r watero r ingress).

~~The first three are the subjects of Coordinated Research Programmes and the last was recently addresse information a n di n exchange meeting.~~

~~IAEA activities in HTGR applications focus on design and evaluation of heat utilization system Japanese th r sfo e HTTR.~~

~~3.1. Co-ordinated Research Programmes (CRPs) in GCR development and application~~

~~3.1.1. CRP on Validation of Safety Related Physics Calculations for Low-enriched GCRs~~

To address core physics issue advancer sfo d gas-cooled reactor designs IAEe ,th A established a CRP on Validation of Safety Related Physics Calculations for Low-enriched GCRs in 1990. At the initiatio statue th experimentaf sP o thif no sCR l datcodd aan e validatio gas-cooler nfo d reactord san the remaining needs were examine IAEdetaie n di th t Ala Specialists Meetin objectivge [RefTh . .1] e filo t l s i gap P validation si CR e ofth n dat physicr afo s methods use corr dfo e desig advancef no d gas- cooled reactors fueled witenrichew hlo d uranium. Countries participatin includ P thign i sCR e China, France, Japan Netherlandse th , , Switzerland Russiae th d ,n an Germany FederationA US e th , .

260 The main activities of the CRP are being carried out by a team of researchers within an international project at the PROTEUS critical experiment facility at the Paul Scherrer Institute, Villigen, Switzerland. Fuel for the experiments was provided by the KFA Research Center, Juelich, Germany, and initial criticality was achieved on July 7, 1992. Experiments are being conducted for graphite moderate systemU dLE s ove rangra experimentaf eo l parameters, suc carbon-to-uranius ha m ratio, core height-to-diameter ratio, and simulated moisture ingress concentration, which have been determine participatine th y db g countrie validatios sa n data needs Paue Th l. Scherrer Instituts eha been highly willing to incorporate experiments as defined by the several participating countries to provide results focuse thein o d r validation data needsmeasurementy Ke . s being performet a d PROTEUS whic providine har g validation data relevan curreno t t t advanced HTGR designe ar s summarize Tablsummarn dA i . e1 PROTEUf yo S condition gives si Tabln ni . e2

~~Tabl : Measuremente1 PROTEUt sa S~~

* Shutdow wortd nro h in core in side reflector * Effect moisturf o s e ingres rangr fo samounf - e o moisturf o t e on reactivity on shutdown rod worth * Critical loadings * Reaction rate ratios (U-235, U-238, Pu-239) * Neutron flux distribution

~~Tabl : PROTEUe2 S Conditions~~

~~pebbl2 UO e fuel * with 16.76% enrichment * Core equivalent diameter = 1.25m * 4 1. o Cort fro D 8 meH/ 0. * C/U-235 from 5 630 to 11 120 * Water simulated by plastic inserts~~

Also data from the uranium fueled criticals at the Japanese VHTRC critical experiment facility on the temperature coefficient (to 200°C) of low enrichment uranium fuel have been provided by JAERI and analyzed by CRP participants. The results show that calculations of the temperature coefficien generalle ar t y accurat withio et n abou percent0 2 t .

~~3.1.2. CRP on Validation of Predictive Methods for Fuel and Fission Product Behaviour in GCRs~~

The experience base for GCR fuel behaviour under accident conditions was reviewed at an IAEA Specialists ValidatioMeetinn o P 199n gCR i Predictivf a no 0d [Refan , .2] e Method Fuer sfo l and Fission Product Behaviour in GCRs was initiated in 1993. Countries participating in this CRP include China, France, Japan, Poland, Germany, the USA and the Russian Federation. Within this CRP, participants are documenting the status of the experimental data base and predictive methods, cooperating in methods verification and validation and will identify and document the additional needs for methods development and experimental validation data.

~~Technical areas being addressed include:~~

* fuel performance during normal operation * fuel performance during accidents (heatup) non-oxidizing conditions oxidizing conditions

261 * fission product behaviour during normal operation behaviour of gaseous and metallic fission products behaviou f plateouo r t * fission product behaviour during accident conditions behaviou gaseouf o r metallid san c fission products re-entrainment of plateout fission product behaviour in reactor building * performance of advanced fuels

~~3.1.3. CRP on Heat Transport and Afterheat Removal for GCRs under Accident Conditions~~

A CRP on Heat Transport and Afterheat Removal for GCRs under Accident Conditions also bega experience 199th n i d 3an e initiatiobass it t ea reviewes n wa IAE n a n dAi Technical Committee Meeting [Ref. 3]. Countries participating in the CRP include China, France, Japan, Germany, the Russiae th US d Aan n Federation objective establiso t Th s i . P thif heo sCR sufficien t experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advanced gas-cooled reactors during accidents. The scope includes experimental and analytical investigation f heao s t transpor naturay b t l convection, conductio thermad nan l radiation withie nth cord reactoan e r vessel d afterheaan , t removal fro e reactormth . Code-to-code d code-toan , - experiment benchmark beine sar g performe verificatior dfo validatiod nan analyticae th f no l methods. Assessment sensitivitief so predictef so d performanc heaf eo t transport system uncertaintieo st y ke n si parameters are also being investigated. Countries are participating in these benchmarks and experimental activities accordin theispecifio gn t ow r c interests. Tabl list e3 benchmarke sth d san cooperatio experimentn ni s included withi CRPe nth .

~~Table 3: Benchmark Exercises and Cooperation in Experiments Included within CRP~~

~~BENCHMARKS COOPERATION IN EXPERIMENTS~~

~~Code-to-code (analyses of heatup accidents) SANA-1 VGM SANA-2 pebble / prism - open topic GT-MHR ai rwate/ r RCCS ope- n topic HTTR « HTR-10 (a)~~

~~Code-to-experiment HTTR RCCS mockup <*> SANA-1 (1) ST-1565 and others being considered~~

~~Code-to-reactor HTTR RCCS (normal operation) Startup/shutdown~~

~~HTR-10 RCCS (normal operation)~~

~~7995 activities~~

~~262 3.1.4. Coordinated Research Programme in HTGR applications~~

To foster international cooperation in HTGR applications the IAEA's Division of Nuclear Power and the Division of Physics and Chemistry have established a CRP on Design and Evaluation of Heat Utilization SystemHige th hr fo Temperaturs e Engineering Test Reactor (HTTR)e Th . ultimate potential offered HTGRs derives from their unique abilit o providt y e hea t higha t - temperatures (e.g., in the range from about 550°C to 1000°C) for endothermic chemical processes aboved an r highlC fo ,° 0 y 85 efficien andt a , t generatio electricitf no y wit turbins hga e technology . [RefHea4] . t from HTGRusee r b productio dfo n ca s f synthesino s and/oga s r hydroged nan methano steam-methany b l e reforming, productio hydrogef no higy nb h temperature electrolysif so steam and by thermochemical splitting of water, production of methanol by steam or hydrogasification oprocesser f fo coal d an , s which demand lower temperatures, suc petroleus ha m refining, seawater desalination, district heating generatiod an , f steano heavr mfo l recover yr sanoi ta dd miningyan . heaimmediate e Ith th f tn i deman t no e s viciniti d reactor e th f yo chemicaa , l heat pipe coule db developed as a high temperature heat transporter.

Several IAEA Member States are concerned about global environmental problems which result from burning fossil fuels. The application of nuclear process heat can make a significant contribution to resolve these problems. In order to select the most promising heat utilization system(s) to be demonstrate HTTRe th t da , some Member States wis cooperato ht desige th evaluatiod n ei nan f no potential HTTR heat utilization systems. Countries participatin includ P thin gi sCR e China, Israel, Germany, Russia, Indonesia processee USA,e Japath Th .d nan s being assesse selectee P dar CR y db participants according to their own national interests depending on status of technology, economic potential, environmental considerations, and other factors.

~~followine Th beine gar g examined:~~

Steam reformin f methango productior efo f hydrogeno methanod nan l CO2 reforming of methane for production of hydrogen and methanol Thermochemical water splittin hydroger gfo n production High temperature electrolysis of steam for hydrogen production Gas turbine for electricity generation Combined coal conversio stead nan m generation

~~In addition, testing of advanced intermediate heat exchangers will be examined.~~

participantP CR e Th collaboratine sar exchanginy gb g existing technical informatioe th n no technology of the heat utilization systems, by developing design concepts and by performing evaluations of candidate systems for potential demonstration with the HTTR.

~~: to e ar P KeCR y e taskth f so~~

a) Define the R&D needs remaining prior to coupling to the HTTR b) Define the goal of the demonstration with the HTTR c) Prepare design concepts for coupling selected systems to the HTTR and perform preliminary safety evaluations, and d) Check licensability of selected systems under Japanese conditions.

Based on evaluations up to now on technology status, the first priority candidate systems to be connecte HTTstea) e (1 th me o dt R ar (and/o methanJ rCO e reforming gas-turbin) syste(2 d man e system r otheFo .r cnadidat eshal D system continuee b l R& e sth brino dt g stag e theth theii meh o t r technology development when they will be considered feasible to be demonstrated at the HTTR.

~~More detailed information is included in a companion paper [Ref. 5].~~

~~263 3.2. Information exchange meetings (1993-1996)~~

~~respons3.2.1R GC . e under accidenta water o r rai l ingress~~

The IAEA Technical Committee Meeting on "Response of Fuel, Fuel Elements and Gas- cooled Reactor Cores under Accidental Air or Water Ingress Conditions" was hosted by the Institute for Nuclear Energy Technology (Tsinghua University, Beijing, China) in October 1993 [Ref. 6]. Som conclusiony eke s fro Technicae mth l Committe summarizee ear followinge th n di .

The response of gas cooled reactors to postulated air and water ingress accidents is highly design dependent and dependent upon the cause and sequence of events involved. Water ingress may be caused by tube ruptures inside the steam generator due to the higher pressure in the secondary loop. The core can only be affected if steam or water is transported from the steam generator to the reactor. Air ingress is possible only after a depressurization accident has already taken place and has to be looked at as an accident with a very low probability.

Considerable experimental data exists regarding behaviour of OCRs under air ingress conditions. These experiments have shown that self sustained reactio f reactono r graphite witr hai doe t occusno r below about 650° abovd Can e this temperature ther windoa s r floei ai wf w o rates: low flows supply insufficient oxidizin d faio removt an l s e reactioga th ge n products, whereas convective coolin higt ga h flows will overcom chemicae eth l heating. Nuclear grade graphit mucs ei h more difficult to burn than coal, coke or charcoal because it has a higher thermal conductivity making it easier to dissipate the heat and because it does not contain impurities which catalyze the oxidation process.

Two serious accidents have occurred which have involved graphite combustion: Windscale (October 1957) and Chernobyl (April 1986). It is important to clearly understand these accident sequences significane th d an , t difference desige th f n thesi nso e reactors, compare gas-cooleo dt d reactors, which use graphite as moderator and either helium or carbon-dioxide as coolant. Windscale was an air cooled, graphite moderated reactor fueled with uranium metal clad in aluminum. The acciden s moswa t t likely triggere rapia y db d rat f increaseo nuclean ei r heating (thas beinwa t g controllea r carriefo t dou d releasWignee th f eo r energy) which caused failuraluminue th f o e m cladding. This expose uraniue dth m metal, whic s extremelhi y r coolantreactiveai e d th an ,o t , resulte uraniua n di m fire, which cause graphite dth e fire. Wate finalls wa r y use cooo dt l dowe nth reactor after other efforts failed. Chernoby watea s rwa l cooled, graphite moderated reactore Th . rapid surge in nuclear power generation at Chernobyl resulted from a series of safety violations and core neutronic instabilities. Eventually liquid nitrogen was used to cool the burning debris. It must emphasizee b d thacooles ga t d reactor coolans sWindscalea n i neither s ai (a te r havus no ) e core neutronic instabilities suc Chernobye thoss ha th f eo l reactor.

Safety examinations of German modular HTR design concepts are addressing even very hypothetical accidents such as the complete rupture of the coaxial hot gas duct. A large scale experiment, called NACOK, is being constructed at the KFA Research Center, Juelich, Germany to measur naturae eth l convectio ingressinf nprovido o t d an er datg ai validatinr afo g theoretical models.

As a part of the safety review of the HTTR, extensive investigations have been carried out by JAERI of that reactor's response to air ingress accidents including rupture of the primary coaxial accidene th ducs hod ga tan tt involvin rupture gth stana f eo d pip eheap attacheto de closurth o dt e reactoe oth f r pressure vessel. Experimenta analyticad an l l investigations have shown that graphite structures would maintain their structural integrity becaus limitee th f eo d amoun f oxygeo t n within the volume of the containment which is available to oxidize graphite. Further, there is no possibility of detonation of the produced gases in the containment. Experimental test results showed that there is a large safety margin in the design of the core support posts.

264 JAER examines Iha response dth HTT e desigth a f Reo o t n basis accident involving rupture of a pipe in the pressurized water cooler. The ingress of water is sensed by the plant protection system instrumentation resultin reacton i g r scra d isolatio mpressurizee an th f no d water cooler. Analyses show thaamoune th t f ingresseo t d wate insufficiens ri resulo t t openinn i t primare th f go y system safety valves, and the auxiliary cooling system rapidly reduces the core temperatures thereby limiting the oxidation of the graphite structures to acceptable levels. Similar investigations have been conducted by INET for design basis accidents of the HTR-10 reactor assuming the rupture of one or two steam generator pipes.

neutronie Th c effect f moisturo s e ingres corn o s e reactivit contron o d wortd yro an le har being examined in Switzerland at the PROTEUS facility. Neutronic effects of water are simulated by inserting polyethylene (CHy rods into the core as this material has essentially the same hydrogen densit waters ya effece f Th increasin.o t g amount f "waterso firss "i increaso t t core eth e reactivity to a maximum due to under moderation of the neutrons under normal conditions, followed by a reactivity decrease as neutron absorption by hydrogen becomes the dominating factor. Further, water addition into the core has the effect of reducing the worth of the shutdown rods. In the experiments to date, these effects have been well predicted, reflecting perhaps the mature state of reactor physics analysis methods.

o ensurT e ultimatth e e a catastrophe-fregoa f o l e nuclear energy technology, additional analyses of extreme hypothetical accident scenarios should be performed and, in parallel, methods for enhancin passive gth e corrosion protectio graphite th f no e fuel element structured san s coul usede db . Experimental activitie Germanyn si , China, Russi Japad aan n have shown that ceramic coatingn sca considerably increas corrosioe eth n resistanc graphitef eo Technicae th t A . l Universit Aachen yi d nan ResearcA theKF h Center Jvilich, Germany successfua , l coating metho bees dha n developed which is a combination of silicon infiltration and slip casting methods to provide a SiC coating on the graphite. Corrosion tests have been conducted simulating accident conditions (massive water and air ingress t temperaturea ) o 1200°Ct s . Future effort e requirear s examino dt behavioue eth e th f o r ceramic coatings especially with neutron irradiation. Activitie t INEsa T have involved forminC gSi coating graphitn so e structure exposiny b s g the meltemo t d silicon. Oxidation experiments have shown very large reductio oxidation i n rate compare uncoateo dt d graphite. Other activitie INEt sa T have shown that additio superfinf no powdeC fuee Si th l o elemenrt t matrix graphite greatly reduces graphite oxidation because SiO2 is formed by SiC-oxygen reaction thereby partly covering and isolating the graphite micropores from further corrosion. Demonstration of the high resistance to coatinoxidatio C water o Si r f graphitn ai ro g o y nb e surfaces including successful test irradiaten so d structures could result in advantages from a public acceptance point of view as well as a technical poin future f vieth o tr wefo desig f HTGRsno .

e closTh e examinatio f experienco n e presente e Technicath o t d l e Committeth o t d le e conclusion that plant safety is not compromised for design basis accidents. Continued efforts to validat predictive eth e methods against experimental dat worthwhilee aar . Protective coating fuer sfo l and graphite components which provide high corrosion resistance should continu developee b o et d and tested as these potentially could provide assurance of safety even for very extreme and hypothetical water or air ingress accident conditions.

~~3.2.2. Development statu modulaf so r HTGR theid san r future role~~

The IAEA Technical Committee Meeting on "Development Status of Modular HTGRs and their Future Role" was hosted by the Netherlands Energy Research Foundation (ECN), Petten (the Netherlands Novembe0 3 ) o t froworksho N 8 m 2 occasioe role EC r th f th e o 199n n th po o 4f n o Modular High Temperature Reactors in the Netherlands, 30 November to 1 December 1994.

e TechnicaTh l Committee Meetin s convenewa g d withi e IAEA'th n s Nuclear Power Programm e recommendatioth n o e e IAEA'th f no s International Working Grou Gas-coolen po d Reactors (IWGGCRs). It was attended by participants from China, France, Germany, Indonesia,

265 Japan Netherlandse th , , Switzerland ,Unite e Russith d daan Sate Americaf so meetine Th . g reviewed nationae th internationad an l l statu activitied s an followin e th f so g topic higr sfo h temperature reactors (HTRs):

* status of national OCR programmes and experience from operation of OCR's * advanced HTR designs and predicted safety and economic performance * future prospect advancer sfo drole HTRf nationath eo d internationasan d an l l organizations in their development

Though considered an advanced type of nuclear power reactor, helium cooled, graphite moderated reactors have been under developmen almosr fo t t forty years. This Technical Committee Meetin s attendewa g y expertb d s from many countrie e nucleath n i sr power communityd an , represented a significant pooling of experience, technology development and aspirations. While the future role of helium cooled reactors cannot be stated with any certainty, this IAEA Technical Committee Meeting brough focuo t majo e sth r technical issues, challenge benefitd san s affecting their future developmen deploymentd an t .

~~3.2.3. 12th Meetin IWGGCf go R~~

e 12tTh h Meetin Internationae th f go l Working Grou Gas-Coolen po d Reactors (IWGGCR) was hosted by the Netherlands Energy Research Foundation (ECN), Petten, the Netherlands on 2 December occasioe 199 th IAE e n 4th o f Ano Technical Committee Meetin "Developmenn go t Status of Modular HTGRs and their Future Role", from 28-30 November 1994 and the ECN workshop on "The Rol Modulaf eo r e Netherlands"HTRth n i s Novembe0 3 , Decembe1 - r r 1994e Th . meeting was attended by representatives from China, France, Germany, the Netherlands, Japan, Switzerland, the United Kingdom, the Russian Federation and the Nuclear Energy Agency of the observerOECy b d Dan s from Indonesi Unitee th d daan States.

~~The IWGGCR welcomed the representative from the Netherlands to the Working Group as newess it t official member.~~

IWGGCe Th R congratulate Japanese dth e Atomic Energy Research Institute (JAERIe th n )o good progress of the construction of the High Temperature Engineering Test Reactor (HTTR) at Oarai. The IWGGCR also congratulated the Institute of Nuclear Energy Technology (INET), Tsinghua University, Beijin stare constructioth f o tn go HTR-1e th f no 0 Test Modul t INETea .

e meetinTh g provide n internationaa d l foru r informatiofo m n exchange between representative Membef so r countries regarding their Gas-Cooled Reactor programmes membere Th . s IWGGCe oth f R strongly felt thapresene th t t international cooperation conducted withi frame nth e of the IWGGCR in the field of gas-cooled reactors is of benefit to their own national programmes and recommended thaAgence th t y continu informatios eit n exchange actiitie cooperativd san e research programmes in gas-cooled reactor development and application.

~~3.2.4. Graphite moderator life cycle technologies~~

Graphit playes eha importann da t moderatoa rol s ea majod an r r structural componenf o t nuclear reactors since the start of atomic energy programmes throughout the world. Currently there mane ar y graphite moderated reactor operation si n which will continu produco et e power until well intnexe oth t century: also ther graphite ear e moderated reactors currently under constructiod nan otherdesige th n si n stage.

~~The last IAEA Specialists Meeting on the status of graphite technology was convened in Tokai-mura, Japa Septemben ni r 1991. Since that time considerable operating experienc bees eha n~~

266 gained materiald an , s developmen testind an t g programs whicinternationaf o e har l interest have been conducted therefors i t I . e considered appropriat internationae th r efo l expertis nucleae th n ei r graphite broughe fielb o dt t togethe exchango t r e technical informatio graphitn no e lifecycle technologies.

IAEAe Th , followin recommendatioe gth Internationae th f no l Working Grou Gas-coolen po d Reactors (IWGGCR), is planning to convene a Specialists Meeting on Graphite Moderator Lifecycle Technologie e Universitth t a s f Batho y , United Kingdom from 25-28 September 1995A . technical tour of an AGR reactor is also foreseen on 28 September, and a tour of the Windscale site is foreseen on 29 September.

purpose meetine Th th exchangf o t eo s gi e informatio statue th graphitf n so no e development, operation o safetd nan y procedure r existinfo s futurd gan e graphite moderated reactors revieo t , w experience on the influence of neutron irradiation and oxidizing conditions on key graphite properties exchango t d an e information usefu decommissioninr fo l g activities meetine Th . plannes gi d within Internationae frame th th f o e l Working Grou Gas-coolen po d Reactors.

It is intended that the programme should involve all topics from the conception of the reactor design throug safe hth e operatio monitorind nan removae core th th o esaft d f go an el disposae th f o l graphite cores at the end of life. The topics to be included are:

* status of national programmes in graphite technology * carbon/carbon composite in-corr sfo e application * core design * core monitoring * codes and standards * graphite fuel element manufacture * graphite property behaviour * irradiation damage mechanisms * radiolytic oxidation * operation and safety procedures for graphite moderated cores * seismic responses of graphite cores

~~3.2.5. Design and development of Gas-cooled Reactors with Closed Cyde Gas Turbine~~

The International Atomic Energy Agenc plannins yi conveno gt Technicaea l Committee Meeting and Workshop on "Design and Development of Gas-cooled Reactors with Closed Cycle Gas Turbines" at the Institute of Nuclear Energy Technology, Tsinghua University, Beijing, China from Octobe0 3 Novembe2 o t r r 1995.

~~meetine Th beins gi g convened withi IAEA' e frame th nth f eo s International Working Group for Gas-cooled Reactors (IWGGCR).~~

purpose meetine Th th provido f t e o s gi opportunite eth revieo yt statue wth desigf so d nan technology development activities for high temperature gas-cooled reactors with closed cycle gas turbines (HTGR-GTs) especialld an , identifo yt y development pathways whic taky ehma advantage of the opportunity for international cooperation on common technology elements.

Recent advance turbomachinern si head yan t exchanger technology provid potentiae eth r fo l a quantum improvement in nuclear power generation economics by use of the HTGR with a closed cycl turbines e ga HTGR-Ge Th . T offers highly efficient generatio electricaf no l powehiga d hran degree of safety based on inherent features and passing systems. Enhanced international cooperation

267 among national OCR programmes in common technology elements, or building blocks, for HTGRs with closed cycle gas turbines, could facilitate their development with overall reduced development costs. In addition to the common elements being addressed currently through IAEA Coordinated Research Programmes, the technical areas in which international cooperation could be beneficial include fabrication technology and qualification of the coated fuel particles, materials development and qualification, and development and testing of turbomachinery, magnetic bearings and heat exchangers.

The first day will consist of paper presentations on national and international activities on gas cooled reactors, and utility interest and economics of HTGR-GTs. This will be followed by two days of Workshop sessions on the following topics for HTGRs with closed cycle gas turbines:

~~a) power conversion b) plant safety c) fissiofued an l n product behaviour d) materials~~

The Workshops will include technical paper presentations and discussions focusing on the status, needs proped an , r development pathway thesn si e technical areas. Reports wil draftee b l n di the Workshops summarizin statue gth developmend san t need especialld san y identifying pathways for international cooperation in development and demonstration in common technology elements. The final day will involve presentations of reports by the Workshop chairmen to the Technical Commitee and discussion of these reports.

~~3.2.6. 13th meetin IWGGCf go R~~

~~The 13th meeting of the IWGGCR will be convened in Spring of 1996 in Vienna. The topic convenee b o t secone fo th rM 199 n di dTC 6 wil selectee b l t thida s meeting.~~

~~3.3. Status report on GCR technology~~

At its 12th meeting the IWGGCR discussed the question of whether a new report on the status of GCR technology in 1995 should be prepared and issued. IAEA as an organization for promoting international cooperatio providinr fo d nan gforua exchangr mfo informatiof eo advancer nfo d nuclear technologies offered coordination and publishing services for such a status report provided member countries of the IWGGCR support such activity and are willing to provide contributions about their national activities.

The last status report has been issued in 1990 and described mainly GCR designs under consideratio 1988/1989n ni thin I .s repor technicatn o emphasi t pu s l desigswa n detail safetd san y features meantime th n I . e program directions have change almosn di l membeal t r statesw Ne . developments have been initiated, others have been terminated.

In the UK significant progress has been made regarding technical performance and consequently economic figure AGRse th Japaf n so I . n constructio HTTe th f nRo test reacto higr rfo h temperature applications has started and is proceeding on schedule. Process heat application possibilities are being prepared in an IAEA CRP. In China the decision to build a 10 MW HTR test reactor has been made and construction has started. The HTR program hi the US has been modified aimindevelopmenw e th no an t s ga di highla f o t y economic desig modulaa f no r HTGR witn ha integrate developmenturbines e dth ga r Fo . realizatiod an t cooperationa n agreemen bees ha t n made wit Russiae hth n Federation Netherlande th n I . desigR sHT n evaluating activities have been launched within the PINK programme. In Germany, governed by strong antinuclear movements, the HTR program has been terminated, but significant know-how is available and HTR-useful R&D activities are going on.

268 Altogether workine th , g group expresse opinios dit n thaprograe th t m redirectione th d an s progress achieve lase th t yearn di s together with very helpful contribution IAEf so A within four CRPs are important and should be described in a new status report for distribution to IAEA member states. suspectes achievementR Iwa t OC developmente w dth ne thad e san th t s trend tendencied san t no e sar sufficiently known in other interested countries. However, a next report describing the present status should also make cleatechnologR rHT thae th t y currentl ystatusD remainR& .a Backgrounn si d dan reasone delath f commerciar o y fo s deploymenR HT l t shoul e includedb d e goalth ,f presen o s t national strategie theid an s r similarities, i.e. keeping ope nvera y potential e futureoptioth r nfo , shoul elaboratede db .

The working group recommended that IAEA should take initiative for the preparation of a next version of a OCR status report. IAEA was willing to prepare an outline of a report for distribution to working group members for review and comments. The finally accepted outline should provide the basis for subsequent contributions of member states. An expanded outline has been prepared e nexTh t . develo o stet s pi pfirsa t draft base inputn do s from Member States. This i s anticipate r earldfo y 1996.

~~3.4. Other form IAEf o s A support~~

Several form IAEf o s A suppor alse tar o availabl Member efo r States intereste gas-coolen di d reactor t havt whicno sbu eo hmajod r development programmes. Upon official request, technical assistanc arrangee b n edevelopinr ca dfo g countrie providinr sfo g expert advice, training, fellowships and special equipment for research. This will assist developing countries to establish the expertise for incorporating advanced gas-cooled reactor technologies into their power generation programmes futuree inth .

~~4. Conclusions~~

Considerable gas-cooled reactor operating experience has been attained through operation of reactors R basiMagnoe th AG c d d concepan ,x an helium-coolef o t d graphite-moderated HTGRs sha been technically proven with the Dragon plant in the UK, the AYR and THTR reactors in Germany and Peach Bottom and Fort St. Vrain in the USA. Construction is well underway on the HTTR engineering test reactor in Japan and completion and operation of the HTTR and its heat utilization system wil majoe b l r milestone gas-coolen i s d reactor developmen developmenn i d an t f nucleao t r process heat applications. Construction of a test module is planned to begin in 1994 in China. Further development efforts are on going in several countries including technology development for HTGRs with gas turbines for highly efficient generation of electricity, and future plants are predicted to attai verna y high degre f safeteo y through relianc inherenn eo t feature passivd an s e systems.

IAEA programmes foster exchange of technical information and encourage cooperative research on gas-cooled reactors. Current IAEA activities focus on safety technology and heat utilization system technology. Especially for advanced reactors with innovative features, international cooperation can play an important role in their development and application.

~~269 REFERENCES~~

~~1. Proceedings of an IAEA Specialists Meeting on "Uncertainties in Physics Calculations for Gas-cooled Reactor Cores", Villigen, Switzerland, May 1990, IWGGCR/24, IAEA Vienna, 1991.~~

~~. 2 Proceeding IAEn a f Aso Specialists Meetin "Behavioun go Gas-coolef ro d Reactor Fuel under Accident Conditions" Ridgek Oa , , USA, November 1990, IWGGCR/25, IAEA, Vienna, 1991.~~

~~3. Proceedings of an IAEA Specialists Meeting on "Decay Heat Removal and Heat Transfer under Norma Accidend lan t Condition OCRs"n si , Juelich, Germany, 1992, IAEA-TECDOC- 757, IAEA, Vienna, 1994.~~

~~4. Proceedings of an IAEA Technical Committee Meeting on "High Temperature Applications of Nuclear Heat", Oarai, Japan, October 1992, IAEA-TECDOC-761, IAEA, Vienna, 1994.~~

5. J. Cleveland and I. Lewkowicz, Status of the IAEA Coordinated Research Programme on Design and Evaluation of Heat Utilization Systems for the HTTR (Presented at the 2nd International Conferenc Multiphasn eo e Flow, Kyoto, Japan, April 1995).

. 6 Proceeding IAEn a f Aso Technical Committee Meetin "Responsn go Fuelf eo , Fuel Elements and Gas-cooled Reactor Cores under Accidental Air or Water Ingress Conditions", Beijing, China, October 1993, IAEA-TECDOC-784, IAEA, Vienna, 1994.

~~270 LIST OF PARTICIPANTS~~

~~Barnert. H , Research Centre Julich (KFA) Institute for Safety Research and Reactor Technology P.O. Box 1913 D-52425 Julich, Germany~~

~~Bastien, D. DMT/DIR CEA/SACLAY 9119 Gif-sur-Yvett- 1 Cedee- x France~~

~~Brey. L , Nuclear Power Technology Development Section Divisio f Nucleano r Power, IAEA 0 10 P.Ox Bo . A-1400 Vienna~~

~~Chen, H. Institut f Nucleaeo r Energy Technology Tsinghua University 100084 Beijing China~~

~~Fox, M. Integrators of System Technology P.O. Box 985355 WATERKLOOF 0145 South Africa~~

~~Gao, Z. Institute of Nuclear Energy Technology Tsinghua University 100084 Beijing China~~

~~Gillet. R . DTP/SECC CEA/GRENOBLE 17, Rue des Martyrs 3805 4Grenobl- eCede- x9 France~~

~~Golovko. F . V , OKB Mechanical Engineering Nizhny Novgorod Russian Federation~~

~~Hayashi, T. Departmen Nucleaf o t r Engineering Tokai University 2-28-4 Tomigaya Shibuyaku Tokyo,Japan~~

~~Van Heek, A. 1 x Bo . O ECN . P , 1755 ZG Petten Netherlands~~

~~271 Kobayashi, O. Fuji Electric Co., Ltd. 1-1 Tanabe Shinden Kawasaki-ku, Kawasaki city Japan~~

~~Kunitomi, K. Japan Atomic Energy Research Institute 3607 Narita-cho, Oarai-machi Higashi-ibaraki-gun, Ibaraki-ken 311-13 Japan~~

~~Matsuo, E. Turbomachinery Laboratory Technical Headquarters Nagasaki Research & Development Center Mitsubishi Heavy Industries, Ltd. 1-1 Akunoura-machi, Nagasaki-shi Nagasaki 850-91, Japan~~

~~Neylan, A. General Atomics P.O. Box 85808 Diegon Sa 92186-978A C , 4 USA~~

~~Sato. K , Advanced Reactor Fuel Division Nuclear Fuel Industry Ltd. 3135-41 Muramatsu, Tokai-mura Naka-gun, Ibaraki 319-11 Japan~~

~~Sun. Y. Institute of Nuclear Energy Technology Tsinghua University 100084 Beijing China~~

~~Tanaka, T. Departmen HTTf o t R Project Japan Atomic Energy Research Institute 3607 Narita-cho, Oarai-machi Higashi-ibaraki-gun Ibaraki-ken, 311-13 Japan~~

~~Tsuchie. Y , Researc Developmenh& t Department The Japan Atomic Power Co. (JAPC) Ohtemachi Building, 1-6-1 Ohtemachi Chiyoda-ku, Tokyo 100, Japan~~

~~Wu, Z. Institut f Nucleaeo r Energy Technology Tsinghua University 100084 Beijing China~~

~~272 Weisbrodt, I. Loner Hohenweg 22 D-52491 Bergisch Gladbach 1 Germany~~

~~Xu . Y , Institut f Nucleaeo r Energy Technology Tsinghua University 100084 Beijing China~~

~~Zhang, Z. Institute of Nuclear Energy Technology Tsinghua University 100084 Beijing China~~

~~Zhong, D. Institute of Nuclear Energy Technology Tsinghua University 100084 Beijing China~~

~~273~~