IAEA-TECDOC-431
LOW-TEMPERATURE NUCLEAR HEAT APPLICATIONS: NUCLEAR POWER PLANTS FOR DISTRICT HEATING
REPORT OF AN ADVISORY GROUP MEETING ON LOW-TEMPERATURE NUCLEAR HEAT APPLICATIONS: NUCLEAR POWER PLANT DISTRICR SFO T HEATING ORGANIZEE TH Y DB INTERNATIONAL ATOMIC ENERGY AGENCY AND HEL PRAGUEDN I , 23-27 JUNE 1986
A TECHNICAL DOCUMENT ISSUED BY THE INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1987 LOW-TEMPERATURE NUCLEAR HEAT APPLICATIONS: NUCLEAR POWER PLANTS FOR DISTRICT HEATING IAEA, VIENNA, 1987 IAEA-TECDOC-431
Printed by the IAEA in Vienna August 1987 PLEASE BE AWARE THAT ALL OF THE MISSING PAGES IN THIS DOCUMENT WERE ORIGINALLY BLANK The IAEA does not normally maintain stocks of reports in this series. However, microfiche copie f thesso e reportobtainee b n sca d from
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Nuclear powe alreads lona ha r r g fo yperio d playe vera d y important productioe th rol n i e electricitf no durind an y g 4 nuclea198639 e th r, power plants in 26 countries, with 270 GW(e; total net capacity generated totae th lf nearl o electricit % 16 y y productio worlde th n .i n This production corresponds to about 560 million tonnes of equivalent coal. If the present growth is maintained and if there are no cancellations and no slow-dow implementatioe th n i n plannef no d projects worldwide th , e nuclear power capacity is expected to be around 370 GW(e) by 1990, with nuclear power contributin world'e th f o s % electricag20 l energy supply. Besides generating electricity, nuclear reactors can also supply heat as a primary energy for heating purposed and for industrial needs. Technical and economic studies in several countries have shown that the heat delivery from nuclear source mann i ys i scase s competitive with fossil-fuelled plantcontributins i d san e sam th e t th a timg o t e cleanlines environmente th f o s . Low-temperature nuclear heat gainea s a d co-product of electricity from nuclear power plants (co-generation)is used already in a number of countries for the supply of warm water or steam from very different type reactorf o svariout a d an ss capacities. Compared with nuclear co-generation plant e specializeth s d nuclear heating plants n earla arn i ye stag developmenf eo d Implementationtan .
The IAEA reflected the needs of its Member States for the exchange of information in the field of nuclear heat application already in the late 1970s. In the early 1980s, some Member States showed their interest in f heao te th us efro m electricity producing nuclear powee th r n planti d an s developmen f nucleao t r heating plants. Accordingly a technica, l committee meeting wit workshoha organizes wa p 198n i d revieo 3t statue th w f so nuclear heat application which confirmed bot e progresth h s mad thin i e s field and the renewed interest of Member States in an active exchange of information about this subject. In 1985 an Advisory Group summarized the Potential of Low-Temperature Nuclear Heat Application; the relevant Technical Document reviewin situatioe gth e IAEA' th n si n Member States swa issued in 1986 (IAEA-ThCDOC-397). Programme plans were made for 1986-88 e IAEans asketh d Awa promoto t d exchange th e f informationo e , with specific emphasi desige th n no s criteria, operating experience, safety requirements and specifications for heat-only reactors, co-generation plant d powean s r plants adapte hear fo dt application. Becausa f eo growing interest of the IAEA's Member States about nuclear heat employment districe ith n t heating domaine n Advisora , y Group meetin s organizewa g d by the IAEA on "Low-Temperature Nuclear beat Application: Nuclear Power Plant Districr fo s t Heating Praguen "i , Czechoslovaki Junn i a e 1986e .Th information gained up to 1986 and discussed during this meeting is embodied in the present Technical Document.
hopes Ii t d that this report would serv e IAEth eA Member Staten a s sa useful technico-economical informatiof o ne us abou e statue th th t f o s low-temperature nuclear heat and give ideas to those who are intending to employ this source of heat as substitution for fossil fuel. EDITORIAL NOTE
preparingIn this materialpress, the International the stafffor of Atomic Energy Agency have mounted and paginated the original manuscripts and given some attention to presentation. The views expressed do not necessarily reflect those of the governments of the Member States or organizations under whose auspices manuscriptsthe were produced. thisin The bookuse of particular designations countriesof territoriesor does implynot any judgement 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 specific companies or of their products or brand names does not imply any endorsement recommendationor IAEA. partthe the of on CONTENTS
1. INTRODUCTION ...... 7 .
2. HEAT SOURCES FOR DISTRICT HEATING ...... 8
2.1. General approach to the selection of heat sources for the DH ...... 8 2.2. Implementatio f nucleano r heat sources ...... 9 .
. 3 NUCLEAR REACTOR DISTRICR SFO T HEATING ...... 2 1 .
4. STATUS OF THE LOW-TEMPERATURE NUCLEAR HEAT APPLICATION IN THE IAEA MEMBER STATES ...... 13
4.1. Canada ...... 13 4.2. China ...... 7 1 . 4.3. Czechoslovakia ...... 20 4.4. Finland ...... 25 4.5. France ...... 8 2 . 4.6. German Democratic Republic ...... 37 4.7. Germany, Federal Republi f ...... o c 1 4 . 4.8. Hungary ...... 47 4.9. Switzerland ...... 8 4 . 4.10. USSR ...... 1 5 .
5. CONCLUSIONS ...... 65
ANNEX 1: MAIN DATA OF NUCLEAR HEATING PLANTS ...... 67
ANNEX 2: USSR SITING REQUIREMENTS FOR NUCLEAR DISTRICT HEATING PLANTS AND FOR NUCLEAR CO-GENERATION WITH REGAR RADIOLOGICAO DT L SAFETY ...... 1 7 .
Lis f Participanto t s ...... 5 7 . 1. INTRODUCTION
In 198s expectea 6- - dnuclea r power reactors worldwide have accumulated 4000 reactor year f operation o snuclea4 39 e rTh powe. r plants f countrie6 198o 2 (NPP d n 6n operatioI i en ) e sth wity GW(e0 b n 27 h ) capacity gav e world'th e f nearlo s% electriy16 c energy.
For more tha year0 3 ne NPP e th playins ar s n Importana g e t th rol n i e safe and reliable generation of electricity. Public confidence, shaken by the Three Mile Island accident, was coming back and until spring 1986 some hesitating countries starte o decidt d n favoui e f nucleao r r powere Th . Chernobyl accident in late April 1986 changed again the situation rapidly: The situation differs from country to country and in some of them it may slow dow e developmenth n f nucleao t r power evey nn ma somi , leat i eo t d the deferring of nuclear programme. Before Chernobyl in the IAEA's calculations the forecasts for nuclear generating capacity was some 40O GW(e) aroun e yeath d r 1990, what should give mor f enucleao tha% 20 n r share in the world electricity generation total. It is conceivable that these figuree affectedb y ma s . Most countries with developed nuclear installations are, of course, well beyond the point of no return from the nuclear programme Directoe Th . r Genera e Internationath f o l l Atomic Energy Agency, Dr. U. Blix, stated in his speech on June 2nd 1986 at the opening session of the ENC'86 Conference in Geneva what measures must now be take n- mainl n joini y t international programmes improving nuclear safet nuclead an y r power plant operation o avoit s d another nuclear acciden f thio t s nature. Eve f I possibln y slowe e foreseeb dn downca nt i , that nuclear power will significantly contribute to the overall energy supplie e futureth n i s.
It is well known that the substitution of nuclear sources for fossil fuelled ones in electricity production is proceeding much faster than in the heat provision area. Still, more than one third of fossil fuel is consumed for heating purposes In the communal as well as in the industrial domaine.
e IAEAn spit th I lit a t f , eo whe n selecting information froe Membeth m r States about their interes r employmenfo t f nucleao t r heat, foun t thaou d t most of the studies or prospective plans are considering the nuclear heat r usdistricfo e t heating. 2. HEAT SOURCES FOR DISTRICT HEATING
The heat needs in industrialized countries are rather high and as already mentioned above severan ,i thef lo m even more tha thire non f o d the fuel currently used is being consumed for heating houses and for supplying industry with process heat. Should these amount fossif so l fuel savee sb y replacinb d g them with nuclear generated heat coult ,i ) d(a bring high savings in the countries' fuel economy, (b) save these products as valuable chemical raws and (c) contribute to the cleanliness of the environment.
In the USSR, where large demands for heat occur, the production of watet stead fore ho ran th heaf mo mn i tconsume 198n i d O more tha0 60 n million tons of equivalent fuel. It is therefore not surprising that this large country with high heat needs startede th alread e us o yt lono gag nuclear substituta hea s ta fossir efo lconstantls i fuel d san y payinn a g even increasing attention to this way of heat provision. They are, however, much smaller countries with a similar policy: as an example Switzerland can be introduced, where according to the " Bundesamt fuer Energiewirtschaft" studies were made confirming the advantage of nuclear heat application substitutin fossir fo g l fuel wit gooha d efficiencd yan economy and with a substantial contribution to the improvement of the envi ronment.
The heat policy of each individual country is usually a complicated tass alreadkwa t andi y s e IAE,a showth A n nTECOÛC-397i waye ,heaf th s o t provision are very much "country-specific". When a decision is taken to build up centralized district heating systems (CDHS), not only the corresponding heat sources are to be developed but all other parts of the systems have to be implemented as well, such as heat feeding network, heat exchanger stations etc.
Only in a few countries with a long tradition of existing CDHS conditions were so positive that further development of larger CLHS became a logical step in the medium and long-term prospective plans. In several other case developmene sth CDHf ts o foun costlo Swa to dd therefor an y e several compromises were accepted in some countries (.outlooking, of course, from political constraints publie ,th c attitude etc.;.
Some general views featurin e philosophth g selectioe th e f th o y f no most suitable technology servin s futurga e heat sourc districr fo e t heating (Drij syste e introducemar e followinth n i d g sub-chapter.
1 Genera2. l approac selectioe th o t h f heao n H tD sourcee th r fo s
Choosing among technology options is one of the most complex questions which decision-makerse e solvehavb th o t ey b d earln a t yA . stage of this decision process some very important preparatory steps are made tb od severa ean l analytical documents prepared e b hav o t e , suc: has
long-term heat consumption forecast long-tere th a par f s o ta s m country's energy policy
- assessmen differenf to t technologies acceptanc publie th y cb e - assessment of the feasibility of the candidates technologies in terms of their availability with respect to the infrastructure, economy, safety, environmental impact etc.
Several computerized evaluation models exist for the energy planning technologe th asr fo wel s yla assessment. This typ planninf eo g studies with the assistance of the IAEA experts were and are being developed for several Agency's Member States.
furthee Foth r r ste decisionf po , i.e choice .th implementinf eo a g proven technology in the district heating system, the so-called TCM methodology (technology choice model) can be used, enabling
structurin probleme th f go , - assessing possible impact of alternatives, - determinin value th g e structured ,an - synthesizin informatioe th g f previouno s step evaluato t s d ean compare alternative actions.
Basically this model balances six important criteria, i.e. environmental impact, healt d safetyhan , socio-economic effects, public attitudes, feasibility and economy, when comparing technologies for the identified prime site.
As said IAEe assistins ,th Awa a lon r g fo g tim Membee th e r Staten i s their case studies related to energy planning and in a similar way it can hel n selectioi p n studie wayr coveo fo st s e country' rth s heat needs, mainly in connection with the nuclear heat application.
2.2 Implementation of nuclear heat sources
In several countries, mainly in countries with high heat needs, the combined production of electric power and heat (co-generation - CHP), now appears to be the most rational method of exploiting fossil as well as nuclear plants. A French comparison of heat generating costs (in French Franc n 1986i s introduces )i Tabln i d . e1
TABL . COMPARISO1 E HEAF O N T GENERATING COSTS (FF/kWh-1986)
Plant Fuel Labour Electric Maintenance Total depreciation cost coat power and coat consumption servicing
PWR, NP 300 typ«' 6,90 2,28 6,61 0,35 1,24 11,38 950 MWth <2) 5,37 1,77 0,47 0,35 0,96 8,92 (1) THERMOS reactor 6,12 2,70 2,85 0,35 1.19 13,21 100 to 2OO MWth
Delivery from EOF power plants, 3OO 11,03 to AGO MWth (1)
Oil-fired plant1' 0,77 8,70 0,36 0,17 1,08 11,08
Coal-fired plant 1,56 12,40 0,48 0.35 2,17 16,96
(1) 4 200 h/year 0 h/yea40 (2S )r Figures in Table 1 show Chat the lowest costs of 1 ktfh(th) are reache hea y b electricitd tan y co-generatio P (PWR-equippedNP e th n ni ) with a long enough yearly operation tiae (5400 h/year). It is well known that under present technico~economic conditions, nuclear power reactors, compared with conventional power units characterizee ,ar highey b d r construction costs and lower costs of fuel. So a nuclear reactor dedicate heao t d t production mus operatee tb fult a d l bas elona loar gfo d period in the year; usually the figure of 4000 h/year is given as a limit. Concerning the fossil fuelled plants, the actual exploitation of fuel combinee foth r d productio reconstructen ni d condensation power plants with regard to producing heat is smaller than in new fossil heating plants. This is reflected mainly in the coefficient of utilizing the production of electricity under the combined method of producing electricit heatd yan , i.e highea n .i r fuel consumption spitn I .f o e this, this metho capabls i d providinf eo necessare th g e b y n heatca r ,o use replaco t d e smaller municipa d industrialan l heating plants, which burn fossil fuel qualityf so lowet ,a r investment costs. f o Thi s si concern mainl countrier fo y sy reason whican r hfo s decideo t t dno implement nuclear heat sources. othee th rn O sid whe~ e nucleana r power source operate lona r gfo d year-perio fula n li d base power loa co-operatdo t mod s eha e with other local peak power source supplyinn i s g centralizea hea o t d heating system, several technico-economic questionsolvede b o t ,e ssucar : has
the achievement of high efficiency in the energy conversion and effective consumptio heaf no t from nuclear power source with optimized lost electric power output replacing fossil fuel nucleay b s r maximua hea o t m possible extent
- optimization of the co-operation of Individual heat sources of the centralized heating system generating lowest heat production costs (optimization of the so-called system and plant factor h)
minimization of investment costs of the centralized heating system.
A nuclear power source cannot be for objective reasons built immediately in the centre of electricity and heat consumption. This means that the heat has to be transported to the place where it is consumed by a system of heat feeders whose efficiency depends again on the required transmitte dannuas it powe d l ran exploitation .
necessare Th y conditio effectivf no e exploitatio classicaf no d lan nuclear condensation power plant producinr sfo existencge th hea s i t f eo centralized heating systems ,enlargee b whic n hca d gradually b y integrating (Interconnecting) regiona d localan l heating systemsn I . founding these systems, the basic investments include the costs of feeding the heat fropowee th m re plac planth f consumptioeo o t d costnan f o s connectin locae gth l centralized heating systems, mai e i.eth n f .heao t feeder and the local heat feeders. These costs are usually shared between the heat supplier and the consumers (municipalities, factories, etc.).
An effective design of the whole system can be achieved by correctly estimating the growth of heat consumption in the consumer network, correct operatioe choicth f eo n characteristic heat-carryine th f so g f mediuo d an m
10 the source, i.e correce .th t choiccoefficiene th y f An eo . ah t inaccuracy in these calculations will be reflected in the effectiveness of the future production more markedl yheae supplies thath i t f ni d from local sources. The attraction of a district heating network, can be shown on its two main characteristics, i.e remote .th e hea tcollective supplth d yan e heat distribution regardS A . firse th s t characteristi feedine e cth th f o g networ differeny kb t sources give sflexibilita followinn i y loae gth d output in time and facilitates to get through outages due to local repairs, maintenance etc. The collective distribution gives the possibilit"centralizee th e us o t y d heat differenn i " t amountn so different network e partth f so . Thus this "size effect" lead loweo t s r costs, because the distribution between various sources concerns also the sharing of the average investment, operation and maintenance changes as wel fues la l changes. Regarding the "installed power" and "connected power", French experience *) show that, "because of the non-simultaneous nature of calls and the reserve energies for start-up, the power necessary to satisfy the real needs is lower than the sum of the subscribed (or connected) powers. tako T e this factor into consideration abatemenn ,a applieds r i t fo 2 :10 2000 flat equivalents, 252 for 60 000. On the other hand, the guarantee of delivery to the consumers implies a diversification and a reserve of means installee ;th d capacit % highe25 s i ry tha instantaneoue nth s real sharepowers i d d,an over several sources. constraina Thi s i s t which weight optimizatioe th n so e installations th f n o effects o tw e Th ., however, compensate eac networks.g h bi othe r fo r "
) See. * Hulst.. J « * "Nuclear district heating - technical and economic aspects of the introduction of a nuclear reactor in a district heating network". Nuclear Energy, 22, No.4, p.251 -9, 1983.
11 . NUCLEA3 R REACTOR DISTRICR SFO T HEATING
There are in principle two main sources of low-temperature nuclear heat, i.e. the proven reactors of NPPs, where heat is decoupled on account of some lost power output (co-generation, CUP; and the specialized nuclear heating plants (NHP). Because a comprehensive description of both these sources from different views is given in the IAEA TECW)C-397, only few additional remarks wil introducee lb d here:
a) Low-temperature nuclear reactors mainly the PWRs (or PhWRs; are widely used or are envisaged for nuclear heat supply in several IAEA Member States like Canada, Czechoslovakia, Federal Republic of Germany, Finland, France, German Democratic Republic, Switzerlan USSRd r an d Fo . similar purposes also other types of nuclear reactors are used. The ChP e takeb a prove n s na ca n technolog botr fo yh warm wate stead an r m supply in various sizes. Some additional informatio relevano t n t chaptere th f so IAEA TECDOC-39 e founb chapten n i d 7ca thif o 4 sr report, introducine th g low-temperature statuth f so e nuclear heat applicatio Agency'e th n ni s Member States.
) b Specialized NHPundee ar s r developmen countriew fe n i tonld e an sth y 2MWt SLOWPOKE ENERGY SYSTEM at Whiteshell in Canada is in operation since o pilotw td 198 an plant6 s AST-50 e nearin0ar g completio USSRe th A .n i n complete overview of nuclear heating plants which are now in operation, in constructio r undeo n r different stage f developmeno s Canadan i t , China, Federal Republi Germanyf co , Finland, France, Sweden, Switzerland an d USSR, is attached to this report as Attachment 1. .
It can be seen that countries developing NHPs are following very specific way f theio s r utilization e USSth R o :s wil l cover with their AS'i-500 high heat demands in areas with high population density, the Canadian SLOWPOKE ENERGY SYbTE s develope Msmaln wa i e lus de mostlth r fo y remote communities. It is;therefore very difficult to transfer these to other countries without adaptation and therefore the utilization of extent ye much o e projectt dito dth no d; s were stil t provelno n well enough to be ready for the international market. For larger units only a few suitable sites can be found in countries with temperate climates and smal t lcompetitiv no unit e ar s e with fossil-fuelled boilers,t i eve f i n can be expected that their unfavourable environmental effects together wit e fossil-fueth h l availability uncertainties will lea o Nrit d P implementation. In the first stage it will be in countries with developed industry, late otherse r th alse expecte n Th i o. d time horizon coule b d just after the year 2000.
12 LOW-TEMPERATUR. STATUE 4 TH F SO E NUCLEAR HEAT APPLICATIOE TH N NI IAEA MEMBER STATES
In this chapter an overview is introduced showing the status of the w temperaturlo e nuclear heat applicatio thosn ni e Agency member States which participate Advisore th t a d y Group meetin Praguen gi r thosFo o . ewh in 1985 took part at the meeting in Wuerenlingen, relevant parts shown in this chapte e supplementrar IAEe th A o TECDOC-39t s a "progres r o 7 s report" for the elapsed year.
4.1 CANADA
There are two main projects for the production of low-temperature heat using nuclear powe progresn i r n Canadpresene i s th t aa t time, i.e. e suppl wateth t ho steaf / a cogeneratio yo I rd vi an m n with electricitn yi CANOU reactor Bruce th et s a Nuclea r Power Developmen Ontarion i t/ 2 d ,an the constructioSLOWPOKt MW 2 e Eth ENERGf no Ye productio SYSTEth r Mfo f no hea d electricittan Manitoban i y e basiTh .c detail thesf so o tw e projects are provided in the IAEA TECLOC-397 "Potential of Low-Temperature Nuclear Heat Applications", Vienna 1986, based on presentations made at the Advisory Group Meeting hel Wuerenlingenn i d , Switzerland, 1985 September 9-13. These details will not be repeated here but
(a) as a supplementary part to the more technological description of the bruce Nuclear Power Development as given in the above IAEA TECDUO-39 "energe 7th y philosophy n "Energa f "o y Centre Concept" based on CANDU reactors will be presented, and
(.b; the updated status of the construction and testing schedule of the 2 MWt SLOWPOKE ENERGY SYSTEM will be introduced.
) Energ(a Energe y th Need d y san Centr e Concept;
A cursory review of energy consumption in ^anada indicates that 5O% to 80% of energy is consumed at a temperature of less than 10O C and 200°C, that 10%-15 consumes i % temperaturet a d s between 200° 300°Cd Can , less tha consume% 5 n temperaturet da s greate r. C Thitha s showU si n30 n in Figure 1. It is therefore theoretically possible for the CANDU 300 power station to provide up to 90% of the total energy demand in Canada. This expectet figur no significantle b s i eo t d y differen mosr fo t t areas worlde oth f .
Practica economid lan c limitation primarily associated with energy distribution, significantly reduces this potential, however careful optimizatio CANDe 0 energth 30 Uf no y syste particular fo m r energy centre applications should satisfy a sizeable remaining market.
In the Energy Centre Concept, one or more CANDU 300 units are the focal point of an energy delivery system. Various energy users are positioned around the power stations such that there is an optimum arrangemen energr fo t y deliver d utilizationan y .
A large number of possibilities exist for combined energy systems consistin electricityf go , high temperature process stea d loweman r temperature steam and/or wate districr rfo t heating. readile Thesb n eca y tailored to the individual requirements of each area.
13 ENEMY CONSUMPTIO CANADNN I A
100 |-
Figur. 1 e
Energy Consumptio Canadn i n a
Figure 2 describes possible energy system based on the CANUU 3OO nuclear steam supply system, servin totae gth l energy needlarga f so e geographic area. Both high grade energy users (industrial and process users of high and intermediate pressure steam> and low grade energy users (agriculture, fish farms, and district heating) are located near the power station. The final energy distribution system must be designed to meet the need eacf so h area, base factorn o d s suc uses ha r requirements, climate, terrain, and the degree of industrial development.
Figure 2.
A Possible Energy System
14 Energy systemdesignee b n energe sca th cateo t dr y fo rdemand e th f so most areas with new industries established within the Energy System area as appropriate mann I .y casefeasibls i t si servo t e e reject heat from "high quality" energy users to "low quality" energy users.
The standard CAND 0 produce30 U s about 1230 MW(th steaf ;o 469t ma 0 kFa(a) pressure. Steam pressur increasee b n ca e 500o t d 0 kPa(ay jb increasing steam generator area, a fairly minor design change. À more significant, but technically feasible design change, incorporating two small steam generators through which the primary coolant passes in series (Figure 3; would permit a portion of the thermal energy to be extracted at even higher pressures energe th exampl690t r f ya o Fo 0.% kPa(a)e40 .
REFERENCE STEAM GENERATOR ARRANGEMENT
HÖH PRESSURE STEAM
LOW PRESSURE STEAM
— FEEOWATER
PMMARY COOLANT
ALTERNATE «TEAM GENERATOR ARRANGEMENT
CAND •TEA« U10 M SUPM.V SYSTEM
Figur . 3 Arrangemene Higr fo th Pressure Steam Production
The energy (steam utilizee b n )wida ca fron CANLe e i d0 th m 30 U variety of ways, and combination of ways. The relationahip between electrical energy production and process heat productio case th e r whenfo nsteae 100 th passes i mf % o d througha turbin d procesean s heat utilize t turbinda e exhaust condition shows i s n in Figur. e4
15 VARIATION DT ElfCTMCAL AHO PflOCESS HEAT OUTPUT WITH TUBBWt BACK PMSSUfll
ClCCTMCU OUTPUT
Figure 4. Electricit Powea v y r
In most energy system applications however steam woul e provideb d t a d varieta f pressureyo meeo t s t user requirements, utilizing both direct and extraction steam supplies. This is shown in Figure 5.
1 SIMm Irom NSP 2 High prMMira M*x procMno t n «me t High primin MMKI w Vf TvfMn* • . 4lntafiMd4»tS . « pranur» ««tractlan
procw« or dtetrtct h*M Figure 5. 7 Low prMwra «IMm to dMrtd twM T. • ExtUHM unio t lt candio«» Energy Delivery Systems f CondwiMr coaUng ««Mr dtectmi«
•Note Mwiy luftwM/gowalor cofnOKwuon pouM«a «r i . IncludinMttG T . n U|>pingN« g lurWn* lunMw«P L . c .«
16 The status of the Canadian Bruce Energy Centre is essentially unchanged frolase th mt report e hea usee meanTh s r b i .t fo o dt t industrial processes rather than district heating. Difficulties are being experienced in attracting energy users to locate at the site, particularly in view of current low world oil prices. Negotations are continuing with a number of potential customers and it is anticipated that a number of firm committments wil forthcominge lb .
(b) 2 MWt SLOWPOKE ENERGY SYSTEM
The current schedule for the construction and testing of the 2 MWt SLOWPOKE ENERGY SYSTE S.E.S/ M t Whiteshel./a shows i l Tabln ni . e2 Constructio facilite th f s beguno wa y Marc n i nw nearin no h s 198i g d 5an completion. The reactor pool and the building which encloses it have been completed and the in-pool components are presently being fabricated and installed.
t S.E.SMW 2 TABL. . SCHEDUL2 E E
A NOVEMBE8 19 R COMMITTED 1985 MARCH CONSTRUCTION STAKU TE 1986 SEPTEMBER FIRST OPERATION 1986 DECEMBER CONFIRM SAFETY CHARACTERISTICS 1987 SEPTEMBER CONFIRM SYSTEM RELIABILITY AND UNATTENDED OPERATION DEMONSTRATE BUILDING HEATING 1988 MARCH DEMONSTRATE ELECTRICITY PRODUCTION
s expecteIi t d thae reactoth t r wil e starte b ln Septembe i p u d r 19b6 wilt I operatee . lb w powe lo t firsa rt a dverifo t t y satisfactory performance of all reactor systems and confirm safety characteristics. Power wil increasee lb d graduall 100%o t y , followe verificatioy b d f no system reliabilit d unattendean y d operation capability.
e reactoTh r wil connectee lb heatine th o t dg system numbea f so f o r the site buildings to demonstrate building heating capability beginning in the winte 1987/88f ro Rankine Th . e cycle engine wil e installelb e dth following yea demonstrato t r e electricity production.
2 CHIN4. A
The People's Republic of China representative attended for the first time the IAEA's meetings dealing with the low-temperature nuclear heat application, therefor e contributioth e n give t onle descriptiono sth y f no the intentions to start activities in nuclear heat applications but also e frame th a challeng f e nucleao sth r fo er power introductioe ar C PR n i n explained.
17 ) Nuclea(a r powe challenga r- e
Nuclear power is now widely acknowledged as a necessity in the People's Republic of China. The reasons for nuclear power Introduction are based on the following facts.*
- Most densely populate betted an d r developed area locatee sar n i d regions with a lack of indigenous energy resources.
Railways are overburdened with coal transport from the country's main coal fields to large industrial- consumers at distances in the range of 1000 - 2000 km. The country's per capita conventional energy resources are rather low, only world'e abou halth e on tf fo s average comparison I . n with countries rich in energy resources like the USSR or the USA, the figure lowee sar r magnitudee e nearlordeth on f y ro yb . e country'Th capitr spe a energy consumptio vers ni y low 198n .I e 5th primary energy consumption per capita was only about 850 kg of equivalent fuel, i.e. about one third of the world's average, and electricity consumption per capita was about 400 Kwh, i.e. one twentiet e thirtieton correspondine o ht th f ho g figure highlf o s y industrialized countries therefors i t I . urgenn ea t tasthir fo ks country to seek alternative energy sources, nuclear in the first place mee o growine ,t tth g energy demands. ) Potentia(b l need low-temperaturf so e nuclear produced heat numbeÀ nucleaf ro r engineer d heasan t consumers believe, thae th t use of nuclear heat is necessary and feasible in the near future in the People's Republi f Chinaco . Argument snucleaf o favourin e us re neath g t are as follows:
Space w temperaturheatinlo d an g e process steam generation account totae th lf o primar % 25 r y fo energy consumptio PROn i n , i.es i .t i higher thae percentagth n e used n electricit,i y generation. According to some studies already mad n thii e expectee sb respecn ca dt i ttha t the nucleadirecf o e tus r hea tlese b wil st leffectivno somn i e e areas tha co-productioe nth nucleaf no r electricit d hean an yi t mitigating the energy shortage problem. Need oversean i s s supplie equipmenf so d engineerintan g services during designin constructiod an g f smalno l nuclear heating reactors are expected to be significantly less than in the case of large and sophisticated electricity generating nuclear plants. Therefore th e
participation of the country's designers, engineers and industry coul largee b d could ran d give also more "self-determinatione th o t " countr developinn i y nuclean gow r energy systems. Sinc late eth e seventies unrestricted suppl t price (a f chea o yl s oi pmuc h lower than those prevailin worle th dn i gmarket bees ;ha n closed thin I .s connection a number of municipalities concerned with pollution problems begae interesteb o nt nuclean i d r district heating. Petrochemical work othed san r large industrial establishments equipped with oil-burning cogeneration plant coverinr fo s g needn si process heat also started to look, for nuclear plants as substitutes for the existing oilburning ones.
18 Peoplee Ith n 'Republis Chinf co existine th a g national fuel cycle industry is well prepared to serve the introduction of nuclear heating reactors. ) Effort(c s made toward introductioe th s nucleaf no r heat reactors
feasibility studie possibln so e applicatio nucleaf no r heat have been made since early eighties. À study on a possible application of a nuclear cogeneration plant in a petrochemical comple completes xwa 1905n di . This stud initiates ywa d unde sponsorshie rth State th ef po Plannin g Commissio d carriey nb an t ou d a joint a petrochemica tasy b k p u forc t ese l comple Nucleaa d xan r Engineering Institute. Several hundred man-years had been spent on this study. Proposal made by the task was to build a 2 x 450 MWt PWR cogeneration plant. Response thio t s s proposal wer encouragint no e s a g the estimated constructio nucleae n th cos f to r cogeneration plan mucs twa h higher than previously expected. • An increased interest in nuclear district heating appeared in some municipalities during the past few years, .such as.' Dalian, l TABLE 3. MAIN DATA OF THE CHINESE 5 MWt NHR ______Type PWR with ______»ligh t boiling Coolant /aodera to r Prlaary circuit typ f dcelgo e n int«gr. circulât Ion MC core teaparatur« 198°C pretaure 4,5 HPa Heating syat«> teaparature 120°C/60°C procure 1,7 HP« Core ht./dla. 690/570 an aaaenbly deaign square nuabec of aaaeabllea 12+4 tuel eleaent cyl. Kw/6 2 1 power density Fuel initiaini l loading 508 Kg (uo,) . JT enenrichmenr • t av. burnup 4600 KWd/t U 19 (d) Conclusive remarks Further developmen f nucleao t r heat applicatio e People'th n i n s Republic of China is not an easy task. The funding problems and questions of economic competitivenes f nucleao s r heat sources with domestic coae ar l of most concern r nucleaFo . r heating plant o makt s a breakthrouge h into the heat markeMW0 20 e constructiot demonstratio- th t 0 10 a f no n plant seems to be unavoidable. To find sufficient financial support will be very difficul e tinteres th eve f i nf potentia o t l users shoul e highb d . To star a programmt e like this, based mainl domestin o y c industry, would mea o brint n g also other related e infrastructurpartth f o s e th n o e corresponding level payin. w Chinno a higg s i ah interese th o t t developmen f nuclea o expectee t b n rca e dpropepower t i th tha o n s i r,t time alse nucleaoth r heat application will find s placnucleait n i e r programme. 4.3 CZECHOSLOVAKIA Nearly 40% of primary energy sources are being invested into the production of heat in the CSSR; the comparable amount of electricity is about 25%. If we also consider that the energetic efficiency in producing hea s roughli t y double tha f producino t g electricit condensatioe th n yi n cycle, the percentage of heat in utility consumption is even larger (about 75%). With a view to the limited reserves of suitable fossil fuels and their progressively Increasing price on world markets, the considerable consumptio f primaro n y energy source o product s e mos th hea ts i t Important factor in analysing heating systems now and in the future. Among other characteristic factors are the size of basic production funds, number of workers, the effect on the environment and especially the amount of investments int e developmenth o f theso t e systems. (a) Centralized heating systems heat is supplied via centralized systems from central sources, i.e. from power plants, heating plants, and by means of decentralized installations from local sources, i.e. buildin housd an g e boiler roomd an s individual heaters usin l kindal g f fuel o powersd an s . The percentag f centralizeo e d heating systems (hereafter CHS) predominates and continues to increase. Centralized supply, the heating industr e narroweth n i y re wordsensth f , o e enables large central heat source combined an s d productio f heao nd electricit an t e usedb y B o t y. centralizing heat output about 10% of the fuel and manpower can be saved, and the environment can be improved. The combined production of electricity and heat means further marked savings of fuel amounting to 15 - 20% as compared to producing heat and electricity separately in the appropriate installations. The advantage of the combined production is in utilizing a larger amount of heat for heating towns, which would otherwise be wasted in the cooling towers of condensation power plants (about 50% of th e fuel)er example th heaFo n a heatini .t n i , g plant, which suppliea s medium-sized district town, more than 10O thousand tons of lignite can be save usiny b d g centralized source combined an s d productio f electricito n y and heat. 20 The characteristi f heao c t supply region s thu i st onl sno y basen o d the quantitative factors mentione dmarkea alsabovet s i oha t d ,bu qualitative aspect. Unie combinatioth y f thesno approache o tw e s i s capabl f providineo ga complet e picturimportance th f eo f heaeo t supply and its most progressive component, the heat for industry, at a time when the fuel and energy budget has become one of the fundamental limiting factordevelopmene th f whole so th ef o tnationa e laclth k economo t e du y of refine d solian d d fuel f qualitso y and n future,i f fossi,o l fueln i s general. The effect of this problem on maintaining an adequate quality of the environment and securing manpower is not negligible either. The heating industry in the CSSR has a tradition of long standing. In some periods it headed the world's technical development in this field. Beginning wit secone hth twentiese d th hal f o f , large centralized heating systems were gradually buil n Praguei t , Ustd Labemfna , Brnd oan elsewhere, of which namely the Brno system had a very high standard in its time. However, their common featur s thaewa t they supplied heat mainlo t y industrial plants, so that their development was restricted by the development of industry. It was only after 1948 that the complex concept of centralized heating systems was fully developed: a number of systems was built which supplied heat to whole agglomerations, and the systems built earlier were expanded. j Hea(b t centralizee sourceth r fo s d heating systems The present status of the development of centralized heating systems in Czechoslovaki characterizes i a coal-burniny b d g heating plants with stea, back-pressurMW m6 6 d an d 5 bleede5 ean , 25 r , turbine12 , b f o s boilers of 50, 75, 125, 150 and 210 t/h, and hot^water boilers of 58 and 5 MWt11 .oldee th Som rf o econdensatio n powerW M plant 5 5 sd witan 2 h3 blocks are gradually being converted into heating plants and, in this connection, some regional heating systems have already beeintt pu no operation (e.g. Opatovice, tiradec Krilov, Pardubice, iComorany,' Most, Litvénov, Chomutov). bees Iha tn basie founth cdf o futur thae on t e e trendth s i s utilizatio f largeno r power plants, classica nucleard lan r supplyin,fo g heat, which is related to the construction of large regional systems of centralized heat supply. This trend require basie th s c e supplie b hea o t t d froe systeth m m power plants first ,a t classical, later nuclear substantiaA . l part will be forme y transib d t heat feeders capabl f transportino e g thermal energy over larger distances, hot-water systems will be preferred, and steam systems wil gradualle lb y converte water t least ho a t expande o r no t ,o d d further. By convertin e condensatioth g n power plant heatino t s g plantse ,w r coashal ou inferiof lo f abl e o lb mak o e t erus quality alsr fo o producing heat, a considerable amount of gaseous, liquid and superior solid fuels will be replaced in small local heat sources, and a substantial amoun f fueo t l wil e savelb f combinemakiny o b d e us g d electricit d heaan yt productio heatine th n i ng industry cycle. 21 An important reason for using condensation power plants to produce markee th hea s di t improvemen environmene th n i t f townto s dependinn go the burning of coal mostly with a high ash content, as well as the smaller demand for manpower. Ihe system power plants, at first classical and later nuclear, will be the basic sources, however, the construction of regional heating system beins i s g assumed, whose graduae basith s si l integratio locae th lf no system f heaso t supply. Abroad, e.g. in the USSR and FRG, there is a marked tendency to integrate local (island meany b heaf S so ) CH t feeders supplied from large powe heatind ran g plants, from waste-heat source industriaf so l sources and refuse incinerating plants. The gradual integration of local heating systems into larger regional heating systems is also being prepared in Czechoslovakia, e.g., in Ostrava, Brno, in the North Bohemian Region and elsewhere. Saving achievee b e combine n th sca y b d d productio heaf no d tan electricity essentiall wayso tw :n i y by building municipal heating plants (as before) convertiny b - g selected existing condensation power planto t s produc en extena hea o t t determine conditione th e y b dth n i s appropriate actual cases. Althoug absolute hth e fuel saving combineo t e sdu d head an t electricity productio heatine th n i ng industr ye same cyclth e e ,th ear great advantage of producing heat in condensation power plants is the replacing of liquid and superior solid fuels which would have to be burnt in towns by inferior fuel burnt in condensation coal-burning power plants, or by uranium fuel burnt in nuclear power plants. e conversioTh intensivo nt e combined electricit head an yt production in condensation power plants will lea significano t d t saving f fueso l which has to be (for physical reasons) reflected in the decrease of specific heat consumption for producing electricity. At the same time, however amoune th , f electricitto y produce condensatioe th y b d n power plants will be decreased. ) Nuclea(c r heat co-generatio P (CHPNP t )na The present and future development of the Czechoslovak power plant is based constructiomainle th n yo nucleaf no r power plants a resul s A f .to this and taking into account the necessity to decrease fuel consumption, s imperativii t e that large aggloerotation CSSe e supplieth Rb n i s d withheat from nuclear sources. Czechoslovakia is one of the countries in whic hconsiderabla e numbe f studiero d projectsan s have been produced with a view to exploiting nuclear sources (nuclear power and heating plants) for the purpose of supplying heat. e incrementTh f heaso t from nuclear sources e planneth r fo d individual five-yea rfollowss a plan e sar : U withiT 0 230 n 198 6199- 0 9 350 TU within 1998 - 1995 U withiÏ O 25iU n 199 6200- 0 which represent yeae th stotaa r y 2000b lJ T hea. 0 t00 suppl8 3 f o y Assuming that nuclear safety precautions will be repsected, this concept will also hav considerabla e e ecological impact. 22 Fro e technicamth l poin f vie o te actua wth l releas e heas th ha tf eo to satisfy the following basic conditions: - firstly, radiation guaranteo safett s ha y e thaheat-carryine th t g medium supplie e consumeth o t de absolutel b r y safe; secondly, the design has to guarantee minimum degradation of the energe productioth n i y transpord an n f heato t ; - thirdly, the technical design has to be simple and displaying dynamic propertie n accordanci s e wit e requirementth h e consumeth f o s r system; - fourthly economie th , c efficienc e heath t f o releasy e desig head nan t transport has to be competitive with other forms of heat supply. The pattern of heat consumption at the intermediate and low-potential level require t swatesuppliee b heasteamd ho e for o f th an t rmo t n i d. The distribution of hot-water heat from nuclear power plants of the WWEK type has practically been solved in Czechoslovakia and approved on a contractua lSovie e basith y b st l existinpartal n i ybuild gan t Czechoslovakia!! nuclear power plants. As already introduced in the IAEA TECDOC-397, the first Czechoslovak nuclear power plants with 440 WWER unit have been equipped with SKODA 220 MM turbines originally designed as purely condensation turbines with uncontrolled steam bleeding only for the regenerative heating of feed water. These turbine e capablar s f supplyino e MW0 g6 J 8~^) (h f t heao t under two-stage heating of the heating water (two heaters each of 4u MWt; from 70 to 150 °C. This method of combined electricity and heat production is used in the Jaslovske Bohunice nuclear power plant (heat feede r ïrnava;fo r thin I .s cas e producee ratith th e f oo d heat outpuo t t the lost power output is 5.08. generatiow Ane f Skodo n M0 U22 a turbine s developewa s r highefo d r heat provision, i.e MW0 9 . t under two-stage heatin f watego ro t fro 0 7 m MW0 12 120°t unded Can r three-stage heatin f watego r - fro150°C 0 m7 . This typ f turbineo e e Oukovan s th i uses n i d y MPP, which wile b l delivering heat for the city of tirno. In this case the ratio of produced heat output to the lost power output will be 7,14. This type of turbines will be also installed at the Mochovce NPP in Czechoslovaki d thean ay wil e als Nordb P l o NP .delivere e e th GDth r Ro fo t d x WWEe Temelith 4 ( R r P 1000Fo NP n e Skod)th a 1000 MW steam turbines beinw no g e developedar ; they will enabl e heath e t productiof o n 590 MWt under two-stage heating of water from 60 - 120° C and 893 MWt under three-stage heatin f wateo g r - fro150 . 0 m6 °C As it can be seen from Table 4, the nuclear heat produced by co-generation (CUP t ;existina constructed gan d nuclear power plantn I s Czechoslovaki e yeath ry b a200 0 should reach about 3000 MWt wit n annuaa h l . supplThiTJ 0 sf abouo 00 ysuppl 8 3 tf heao y t froe nucleath m r sources would represent a substitution of about l 6OO thousand tonnes of reference e combine th a savin fue r yeao d t pe lan n fue e ri gd du productiol f o n electricity and heat of 730 thousand tonnes of reference fuel. 23 Table 4. NUCLEAR HEAT PRODUCED BY CO-GENERATION (CHP) AT EXISTIN CONSTRUCTED GAN CZECHOSLOVAKIN I P DNP A E YEABTH YR 2000 NPP UNITS HEAT OUTPUT HEAT FEEDERS SUBSTIT.REF. MW(t) TO LOCALITIES FUEL (lO^ton) EBO-1 *) 2 x WWER 240 LEOPOLDOV , 0,145 400 HLOHOVEC EBO-2 2 x WWER 240 TRNAVA 0,102 400 EDU WWEx 4 R 680 BRNO 0,332 AGO EMO 4 x WWER 640 NITRA, 0,256 400 LEVICE ETE WWEx 4 R 365 CESKE 0,264 1000 BUDEJOVICE EKE WWEx 2 R 460 E C I S KO 0,501 1000 PRESOV *) Abbreviations: EB JaslovskOP .NP . . é Bohunice EDU .. . NPP Dukovany , . O EM MochovcP NP . e ET TemeliEP NP .. . n EKE .. . NPP Kosice (d> Nuclear heating plants The problems of utilizing nuclear power sources for heating which have not been solved yet are those related to steam consumption from nuclear power plants and the operation of nuclear heating plants. The possibility of constructing nuclear heating plants is being studied in Czechoslovakia in terms of territorial units which are outside the technica d economilan c reac f long-rango h e heat supplies from nuclear power plants or classical sources; in this connection, the utilization of the 0 typteste50 eT nucleadAS blocke th rf so heatin g plan beins i t g considered in the first place, and nuclear heating plants of smaller performanc seconde th eMWt0 n (1020 i J . 0- These problems are the subject of the "Agreement and program of co-operatio membee th f rno countrie CMEe th n scientificAi f so , technical and design project fiele th f nuclea do n i s r power heat plant d nucleaan s r heating plant producinr fo s g industria heatinr l fo stea d gan m systems." The economic comparison of nuclear heat sources, i.e. nuclear power and heat plants and nuclear heating plants, with the other possible methods of supplying heat has to be carried out individually, from case to case, because the conditions in the individual areas differ. Therefore, the developmental heating-system studies of the appropriate localities (.Brno, Bratislava, Ostrava üudejovice. ,C , Trnava, Leopoldo v- Hlohovec ; always contain economic comparison variante th f so f heatin o s g systems 24 based on the further development of decentralized heating systems founded on the expansion of the existing structure of the fuel base, i.e. primarily on the increase in the share of earth gas, mazut and the ChS variant wit hfossia l source, wit hnucleaa r powe head ran t plant alsr ,o o with a nuclear heating plant. In general, the converted costs of combined electricity and heat productio nucleaa n i n r power plan e alwaytar s lower thane thosth f eo heat-system variants based on earth gas or mazut. Although water is a relatively slow carrie thermaf ro l energy watet ,ho r seems t present,a o ,t e onlbth ey available medium heae ,th t lossee feederth n i ss over distance beinm k f abous o 0 g 10 tabou . 3% t The economic analysis of nuclear heating plants is problematic at present, becaus basie eth c input data suc s investmenha t costs, operational costs, etc. are not available. Available calculations are base n specialiso d t estimates founde derivinn o d nucleae price th th gf eo r heating plant from the price of the nuclear equipment manufactured for the WWER 440 blocks. These calculations have shown that a nuclear heating plant is only justified if heat cannot be supplied in another way, i.e. because of the lack of earth gas, because of the large distance from a fossil power plan nuclear o t r power plant d unde ,an conditioe th r n that the nuclear heating plant will replac considerablea e amoun eartf o t s hga d liquian dlocalite th fue t la n question i y . 4.4 FINLAND e districTh t heating technolog s highli y y develope mors ha e d an d than 40 years tradition in Finland, txisting networks in the towns and communities are on the other hand mostly too small for being adapted to larger-scale nuclear district heating n exceptioA . e Helsinkth s i n i metropolitan area comprising neighbouring cities of Vantaa and Espo and, of course maie ,th n cit f Helsinkiyo . The total energy forecast for the Helsinki metropolitan area by the yea rcorrespondine 2U1th s show i d 0 Tabln an i n . 5 e g figure alse ar s o plotted in Figure 6. TABLE 5. DEVELOPMENT OF DISTRICT HEATING ENERGY IN THE HELSINKI METROPOLITAN AREA Year Population Building Specific Extent of Energy Capacity • tock consumption connect Ion« 1,000 Inhab. 10» «3 Mfh/«3, a \ GWh HH 1982 794 180 SO 72 6,908 2,586 1990 839 211 46 81 8,438 3,392 2000 860 241 42 87 9,454 3,985 2010 250 40 88 9,476 4,112 25 TWh Eiwqy 1980 199O Figur . 6 ePrognosi f Districo s t HeaElectricitd an t y Deman n Helsinki d i Metropolitan Area e comparisoth r fo f differeno n t energy procurement alternativese th , demand forecasts of electricity and district heat for the period 1982 - 2000 were used as the basis. According to these forecasts, the demand for district heat will increase from the 6,910 GWh in 1982 to 9,450 GWh by the year 2000 e correspondinth , g Increas n electriciti e y demand being from 3,900 ÜWh to 6,920 UVh. The energy procurement alternatives studied were the following: coal-fired back-pressure power plant (100/180 MW and 2 x 250/400 MW) - nuclear co-generation power plant (Kopparnäs 40 km pipeline, 2 x 800/90; 0MW nuclear heating plant (SECURE 2 x 400 MW) - nuclear co-generation power plant (Loviisa, 75 Km pipeline, 900 MW) coal-fired co-generation power plant (Inkoo, 55 km pipeline, 800 MW) coal-fire t wateho d r boilers (840 MW) natural gas-fire s turbinga d e power planx 5U/9 7 ( 1t (135) MW). 26 Some conclusions based primarily on economical aspects can be drawn resulte comparativbasie e th oth nth f sf o so e study: - The implementation of 2 new, rather big power plants or heating plants wil necessare lb earle th yn i y1990 s - Nuclear heating plants proved to be the most economical alternative out of tnose studied Heat transmission from Inko d Loviisaoan coal-fired ,an watet ho d r boiler competitivee sar difference th d ,an e compared wit nucleae hth r heating plant is not significant Kopparnäe Th - s nuclear power plant alternativ mors i e e economical than coal, lowes e excepth r t fo tcoa l prices The general rise in fuel prices adds to the economy of the nuclear-based alternatives. At low fuel prices the competitiveness of the Inkoo and the hot water boiler alternatives improves compared with the nuclear-based alternatives A reduction in the required interest of investments in the Helsinki metropolitan area increase econom e K.opparnae sth th f yo s alternative in particular. A rise in the required interest adds to the economy e alternativeoth f spurchase baseth n o d electricityf eo . A feasibility stud carries ywa durint ou d g 198 3198- cooperation 4i n between Imatra ASKA-ATOB A n d Voiman y aMO concerning utilizatioe th f no SECURE nuclear district heating plant for the Helsinki metropolitan area. Considerable attentio alss nwa o pai licensabilito t d sitd yan e survef yo the plant. To be able to start this work it was necessary to elaborate new design criteria, based on the Finnish General ties ig n Criteria for Nuclear Power Plants but amended with a view to the unique features of the SECURE NÜP. Accordin resulte th rathea o f t go s r deepgoing analysi e safetth f yo s features of the SECURE reactor during the feasibility study, the inherent regardee b o t e plansafet s vers th a d ha tf y o y high unsurpassablo N . e demands for improvements or changes in the plant construction can really be expecte resule a potentiath f s o ta d l safety evaluatioe th y b n Authority. Some risks remain, however, possibly leadin furtheo t g r discussion from the point of view of the Authority. They include external events like earth-quakes, airplane crashes, floods, etc., depending on the site. Sabotage and, finally, potential acts of war must also be considered. These two factors are difficult to appraise. However, all these factors have been taken into account in principle in the reactor design. Durin e feasibilitth g y stud f SECURo y Ea preliminar y site selection study was also made. The objectives of this study were the following: - obtainin firse th g t site-related cost approximations, 27 analyzin totalite th g f SËCIAE-relateo y d siting factord san developing adequate site selection methodologies, and obtaining the first reference site to be used for prospective contacts with regulatory bodies, administrative organ f citieso r so citizen groups. e studt includ mads desTh a no wa ys d y fielke a di ean stud dd an y investigations; neither were safety authorities contacted in the course stude oth f y process. The main input data for the site selection process were; the construction of 2 x 400 MW units in close sequency, the sharing of heat supplEspoo r r Helsinkifo fo W y,W M M accordin 0 0 mode,18 e 30 th f lo o t g Vantaar fo W consideratioe M , th 0 12 d outen a an f rno radiu m abouf k so 0 4 t foeconomicalln a r y feasible siting. Three sites,out of the total 43, were submitted to a detailed scrutiny whicn ,i h various aspectfiele th engineeringf o dn i s , economy, ecology, land use, social development, health, and safety were considered. AS a conclusion the working group recommended one site in the northeastern part of the Helsinki metropolitan area to be used as a reference site in later phases of the SECURE study. This site is about 15 km from the centre of Helsinki and about 6 km from the nearest point of the Helsinki district heating network. Nevertheless, the number of population, being abou 0 inhabitantt40 s withi from e k th m5 nradiua 1. f so site, represents quite well the average of the whole southern Finland. 5 FRANC4. E Simila otheo t r r highly industrialized countries, Francs i e developing both main types of nuclear heat sources suitable for district heating applications, i.e e co-generatioth . n systems (CHP) providing electricity and heat from large KPPs as well as the nuclear heating plants (NHP) base specializen o d d "heating-only" reactors. The well known PWR highly standardized French NPPs are all built so that provisio s alwayi n e potentias th mad r fo e l deliverin f thermao g l energy in the form of live steam or pressurized hot water. Two reactors are under development of the "heating-only" type, i.e.: reactorS ThCA e a pressurize, d water reactor derived from ship propulsion reactors, delivering steam at 26b°C and 53 bar pressure; such reactors have a thermal power of approximately 1000 MW. The THERMOS type reactors, derived from "swimming pool" reactors, which deliver watea temperatur o t 130°Ct a r 0 12 ; f thermao e l power of these reactors ranges from 100 to 200 MW. 28 ) Combine(a d productio hea f no d electri tan c power (CHP;. Optimization of the production and operation mode The PWK nuclear development programme started in 1970 in France comprises: - 34 900 MWe units, of which 31 were coupled to the system at the beginnin 1985f go ; - 20 1300 MWe units, of which two were coupled to the system at the beginnin 1985f go . 1308 1 unitW 0d M an se werMW Thre 0 e 90 eunde r constructioe th f o s na start of 1985. This programme has been conducted so as to standardize as far as possibl n seriei e f identicaso l units maie ,th n buildings being builn o t the same plans, housin same th ge unit f equipmento s . The characteristic values for steam at the outlet of the steam generator steae th mf o ssuppl y system are: unitsW M 0 90 Z70°;d r an Cfo r 5 5ba 69 bar and 285°C for 1 300 MW units. (1; Systems for heat withdrawal from the lind circuit: Each nuclear uni equippes i t d wit ha stea m transformer d whicfe s hi with live steam withdraw steae th mt na generator outlet suppliet :i s steam e auxiliartth o e powey th systemuni e rd th stationtan f o s . The characteristics of the steam supplied by each auxiliary steam transformer are: unitsW M 0 35 ;90 188° - tons/hour r Cba fo 2 1 r- unitsW M 0 .30 5 5188°1 - tons/hour r Cba fo 2 1 r- Each steam supply transforme a plant' f o r s auxiliary steam circuit produces roughly 120 000 tons of steam per year to provide for the needs of the power station itself. In total, for the nuclear units in service, this corresponds to approximately 2 500 OOÜ MWh per year, i.e. the combustio l equivalen oi tonO f nearlf OU o no s 0 f fossi25 yo t l fuel. n additioe i station neede th th f o se thermao t nth , l powe f thio r s auxiliary steam supply system can supply approximately 10 tons per hour of steam outside the site, with a rate of availability depending on the number of units and their operating conditions. When heat is supplied outside the power station, a steam transformer or heat exchanger is fitted between the auxiliary steam supply system and the external heat system to forestall any risk of contamination from the section of the auxiliary steam supply system located in the Kadwaste system building (Fo maie rth n schem e Figurse e e 7.;. 29 Figure 7. Heat Withdrawn from Nuclear Power Stations Measures are taken to withdraw steam at the steam generator outlet so as to supply approximately 3OU MW of thermal power for a 900 MW unit and 400 MW of thermal power for a 1 300 MW unit. qualit e e lighth th f tn o yI U feedwate S requirement e th r f o rfo t sse nuclear steam supply systems ,heaa t exchange needes i r d betwee steae th n m supply system of the turbine and the outside heating network. This exchanger places an additional shield between the nuclear fuel and the off-site location. Given the characteristics of the steam produced, these heat exchangers can supply steam at a pressure of up to 40 bar, hot water, or superheate dtemperatura wateo t p ru 250°Cf eo . Large quantities of heat can be withdrawn, if this is done regularly durin operatioe th gpowee th rf n o station, correspondin: to g unitW M equivalenl 0 0 oi ton; 90 00 f so a 0 year r 18 pe tfo r 270 000 tons of oil equivalent per year for a 1 300 MW unit. By reducing the rate of steam flow to the turbine, this withdrawal of steam involve sreductioa electricitn ni y output. Should the heat specifications be low enough, e.g. water at 15U°C or 110°C, an auxiliary turbo-generating set can be placed between the extraction point at steam generator outlet and the heat exchanger. fart steae oth f m expansio thes i n n recovere r electricitfo d y production, diminishing the overall reduction in electrical power output. 30 The measures adopted leave opedecisioa date th f neo withdrao t n w characteristicchoice e th th heaf d eo an t s taken into accoun thir fo ts heat; actually decisioe take,e th b n nn ca afte r commissionine th f o g power station. These measures provide great flexibility regardine th g choice of heat characteristics since there is no relationship between these characteristics and the extraction levels defined for the main turbine auxiliare th ; y back-pressure designee generatinb n a ca s a dt se g function of heat requirements. Even if the main turbo-generating set operates at partial or even very low load, this does not reduce heat supply capacity. These measures do away with any need to execute work on the main series-built turbines, and limit the number of changes to be made to the installations fitted in the turbine hall in the event of a decision taken after construction of the unit has begun. They do not alter the general condition f operatioso unite th .f no (2; Heat extraction from turbines Studies were conducted to define the volume of steam that could be withdrawn fro extractioe mth n point f turbineso produco t sr o t eho superheated water. For limited demand, this type of heat extraction results in a lower los electricitf so y outpusmallea d tan r investment. A heat exchanger shoul placee b d d betwee secondare th n y steam supply e circui systefluie turbine th th th f dd mo f t o carryinean g heat distributed outside the site. All these studies have been so conducted as not to bring into question either the main units of equipment or their operating .conditions. Wit existence hth severaf eo l unita powe n i sr station, potential reserve largee sar . Were larger heat supply contract concludede b o t s , studies woule db undertaken to adapt the design of the power station to the optimum overall leve electricitf lo d heayan t production combined. Howevert ,no this sha yet occurred. (3; Supplying heat outside the power station Technica d economilan c studiescope r usinth fo ef so g heat produce y poweb d r station districr fo s t heatin existind gan potentiar go l industria agriculturar o l l applications have been conducte nuclea8 1 n o d r sites housing units in service or under construction. The studies are up-dated to follow the trend of fossil fuel prices. For economic reasons and because power stations are installed away from urban centres, studies undertaken at each power station for the suppl f hea yo districr fo t t heatin r industriago l applications a hav t eno yet identified any project whose economic value was enough to justify its implementation. Notwithstanding, some studie e undear s r way, e.g stude .th y covering the suppl e tows surrounding f th heaf Troyeit o y no o t td e san th y b s 31 nuclear power statio f Nogent-sur-Seino n e MW0 (tw30 e unitso1 , commissioning schedule r 1988fo d , distanc km;5 5 e . If t sufficiendemanno s i d t withi na reasonabl e distance froe mth power station, the installation of new customers may be envisaged in its vicinity: this meets the concern to develop local activities after construction of the site has been completed. In this perspective of developmen industriesw ne f o t , calorific energattractivn a t ya e prics i e a useful incentive for potential customers. In this regard, the 900 MWe nuclear units ßl and B2 of Chinon, commissione n 190i d 198d 3e an fitte 4ar d with steam transformers, which can eac e industria- 205° hth r supplo ba Ct 7 ton5 1 r hou3 y t lpe sa r estate developed on the Northern boundary of the site of the power station. Negotiations are under way for the installation of several industrial establishments in this estate. ) Nuclea(b r "neating-only" reactors (NtiPj As said, two main types of reactors for NHPs are under development in France: (1) ïhe CAS reactor CAS pressurized water reactors have been developed by the Commissaria à l'Energit e Atomiqu s subsidiarit d an e y TECHNICATOME. CAS reactors are compact and self contained. The CAS (Compact Advanced System) nuclear steam supply system with therma lHW0 powe95 , supplief o d 268°Cr an r .sba 3 stea5 t ma It could be used for: - production of steam only - combined heat t water(stea ho powed r man o j r production using a back-pressure turbo-generatint se g - electricity generation only (300 MWe; using a condensing turbo-generator set. The e schemsteath mf o e supply systen ca s mi show t i n s Figuri n A . b e be seen steao , tw eacm f o hgenerator s i connectes e vesseth a y o t b dl short straight conduc n whici t e primarth h y coolant circulates concentrically back and forth. The primary coolant pumps are integrated wateG S re th boxes n i . This design is based on ÏECUNICATOM's experience in this field (13 PWR in service or under construction; a prototype unit has been in operatio r ninfo n e year t Cadaräche)a s . The compact geometry of the primary circuit makes it possible to reduce the volume of the metallica containment and to simplify the on-site assembly of the primary circuit. 32 Figur. 8 e Nuclear Steam Supply System The containment of the primary circuit consists of a metallica cylindrical central part, closed at each end by a spherical dome (see Figur) 9. e 1 Reactor vessel 2 See«» generator 3 Pressuriser 4 Fuel loading Machine 5 Transfer fuel pool 6 Residual heat rénovai punp rooB 7 Primary pumps handling room 8 Personnel «tr lock Figure 9. HousinContainmene th f o g t Vessel 33 la addition, the nuclear safety auxiliary systems and the fuel storag el protecte poolal e ,ar d from external aggressioa y nb semi-circular concrete tunnel (Figure 10.) Figure 10. Nuclear Island The fuel assemblies are identical (apart from variations in length; to the standard assemblies used in the other PWK power stations built by FKAMAÏGMK. The technology of the main components is that adopted by FKAhATOMÜ levele MW n 0 0 MWe/si 50 00 13 1 d an fo re unitMW 0 equipmenf s90 o e th f to France under EUF's construction programme. Load variation effectee sar d onl controactine e b yth n go l rodf so the reactor. The manufacturing and operation characteristics of this nuclear steam supply systeenvisagee b o t m e n largallowi dus s e it surba n heat distribution networks. projecA supplyinr fo t districe th g t heating networ Parin ki s comprises: - a 950 MWt CAS - nuclear steam supply system - a JO wwe back-pressure turbo-generating set fed by the secondary steam circuit of the nuclear steam supply system - a group of exchangers, connected to the exhaust of this set whic n supplhca y 12ÜO tons/ superheatef o h bar5 2 d- . 0 stea2 t ma 34 The distance over which steam is to be transported is 1? km. The district heating networ f Pario k s require millio8 s n tonf o s stea r yearmpe , which correspond o approximatelt s e e fourtth yon f o h city's heating needs. The needs of the district heating network at base load are presently met as follows: - 36 per cent by heat recovery from the incineration of household refuse r cen supplyinpe y b t 1 1 g heat produce e coal-fireth y b d d electrical power station of Citry-sur-Seine r cen coal-firey pe b t 0 2 d - boilers commissioned ovee lasw th rfe t years. Studies are in progress to replace a large share of the heavy-fuel oil, whic s stil i hInstalo t usen i d l natura y an b ,a fe l s wga electrical l boilers for steam production in summer. (2) e THERMOTh S reactor The THERMOS Nuclear e productioBoileth r t watefo rho s f wa ro n designed by TEŒNICATOHL society, CEA's subsidiary (see Figure 11;. I Cuv« Valid t wattHo r 10 Iafrl|4ran Air cooler It Icbangiuri pria»tr«a Primary beat Exchanger« v Figure 11. Thermos Reactor 12 tchanfeun secondaire* Secondary boat excbanyerl 35 ThEKMUS is basically a slightly pressurized (ca 10 bar) light water reactor using low-enriched uranium, installae bottoe poola th f t o mda . The pool water provides a biological shielding, a confinement barrier by condensatio f primarno y coolant leakag watea d rean reserv corr efo e y incidentcoolinevene an th f to n i g . e reactoTh designes i r ratea r dfo d outpuo t W t terminalM ta 0 10 f so fore approximatel watef th mo n i r W heateM 0 80°t 20 yda 130°C o Ct . Other options are also possible. The whole unit is located inside a double containment of prestressed concrete e inneTh .r pressure vessel preserve from releas fissiof eo n product accidentan i s l conditions e otheTh . r containment ensures protection against hazards of external origin such as explosions, airplane crashes, etc. (see Figure 12.) Figur . Thermo12 e s Process Flow Sheet The THEKMOS reacto a typica s ri l "three-circuit" reacto d thiran s fact together wit s intrinsiit h c safety characteristics allows sit installatio denseln i n y populated urban areas. The main application of the ÏHKKMOS reactor is therefore the district heating s characteristic,it e suser seawateb enablfo do t t i er desalination as well. o possibilitieTw THEKMOe th f so S reactor application were examined, heae MW tdistricU e suppl1U th e r i.eth fo y t. heatin e citf th yo f go Grenoble and 150 I-tWt supply to several heating networks of the South-Uest suburb of faris. In the case of Grenoble, its network needs 650 000 MWh per year, but since 17 per cent of the needs are covered by the recovery of heat from the household refuse incineration plant and coal is available from a neighbouring mine, the utilisation time of the nuclear boiler is reduced to 2 500 hours, full power equivalent per year, the economic value of this project is not sufficient today to make it viable. 36 The other project (South-West suburb of Paris; was abandoned because of the overwhelming Initial Investment required to Interconnect the separate networks. Studies are in progress for other, possibly more Interesting applications, especially outside France. 6 4. GERMAN DEMOCRATIC REPUBLIC 4.6.1 Fundamentals Since 1981 factorie e Germaworkd th an sn i sn democratic Republic have limited the use of solid and liquid fuels for heat supply. To accomplish the task imposed on them they concentrated their efforts especially on applying energy as efficiently as possible, - substituting brown coar soli liquid fo l dan d fueld an s - utilizing more intensively waste treatmen e co-generatioth d an t n technology. In the application of the nuclear heat gained by co-generation from a NPPe Nucleath , r Power Plant "bruno Leuschner" Greifswald together with industrial enterprise scientifid an s c Institutions carriet ou d investigation e heath tn o ssupple nearb th o t yy tow f Greifswaldo n . Within a very short time a project was worked out and after a construction time of only one year this supply started with a thermal power of 50 MW on 27 December 1983. The hot water of 180°C (2.86 MPa) is transported from e exchangeth o t P rNP statioe e towth th nf no Greifswal d throug ha transi t pipeline wit a hdiamete a lengt 0 md e m80 an f Th morf ho o r . e km tha 0 2 n exchanger station in Greifswald is the distribution center of hot water int e networ reture e townoth th Th f .no k temperatur e transith n i e t pipeline which is now 80°C, will be decreased to 60°C when the corresponding part of the heat will be used for heating a greenhouse. The e schemheath tf o esuppl y syste furthed man r e detaile founth b n n i d ca s following paragraph 4.6.2 where the District Heating of the town Greifswals from the NPP "Bruno Leuschner" is described. e ProjecIth n t Requirements were realise e Nationath f o d l tioarr fo d Nuclear Safety and Radiation Protection of the GLiR for preventing release of radioactivity into the heat distribution network from the heat extraction system at the NPP. Full text of the Requirements can be found e "Nucleath s Annea n i 2 xr Heat Application, Proceeding a Technica f o s l Committee Meeting and Workshop, Cracow 5-9 December 1983", page 409, IAEA Vienna, 1984, SÏI/PUU/679. Successfully results were already receive a grea d an dt deaf o l experienc s gainewa e d with heat extractio "BrunP nNP ofro e Leuschnermth " 0 e MWfutur15 Th . o uet p deman f abouo d 0 MW30 t wil e realiseb l y heab d t extraction fro l unitmal f thio s s hPP. The experience froe describemth d first nuclear heat supply systes mi confirmin e technicath g economid an l c applicabilit f thiyo s co-generation technology on further NPPs in the GDR. Assessments are also being made investigating possible implementation of the Soviet AST-heating reactors e OUR th t concret bu n , i t eexist ye project t .no o d s 37 4.6.2 District Heatin e Towth nf go Greifswal d froe Nucleamth r Power Plant "Bruno Leuschner" (a) Development of the heating requirements of the town Greifswald e latth eUo t p70' e heaath t demand e towth nf o sUreifswal d were covered by oil, hard coal and brown coal. These demands derived from 3580 heating-degree days and 241 heating days, based on an average room temperature of 20°C and an average outdoor temperature of 12°C as heating boundary. The prospective demand will amount to 300 MW. The structure of the consumers consists of the municipal area coverin r cenpe t5 5 g industrial area covering 20 per cent agricultural area covering 17 per cent. This means that finally 9000 dwellings, most of the industrial plants and a greenhouse will be supplied with heat. AS can be seen, the greater demand for heat arises from the municipal area and depends essentially on the outdoor temperature. Due to the climatic influence of the Baltic Sea there is a slight rise in temperature in the months of October to December an a decreasd e from Januar o Mat y rh compare d wit e averag e hth GLR th n i .e The graphs of the daily heating requirements during the heating period show only little variations. A typical depression at night cannot be distinguished. (bj Technical projec f co-generatioo t n In elaboratin e technicagth l projec f nuclear-produceo t d heat via-co-generatio e followinth n e gtake b aspecto nt intd ha so account: - The project has to be an example of economic heat supply from a nuclear power plant. - À decrease in the electrical power of the nuclear power plant in the maximum power demand times has to be excluded. The maximum extractio f thermao n M5 r 7 turbineWpe l e b powe y ma .r To ensur a stable e heat casy supplan e n thermai y e b lo t powe s ha r taken from all units. In consequenc f thio e s hot-water heatin s effectei g n threi d e steps via the taps 3, 4 and 5 of the turbines. The limit of 75riW results from the permissible speed of the steam in the taps. The producer of the turbine K-220-44 permits speed m/sÜ 6 r shor s,fo o t fro t0 m5 period s even 0 m/so7 f . e seeb nn Aca fros m e Figurfixeth 3 d1 e extractio f steao n m enables t wateho o react e r th ha maximu m temperatur f a pressur180°o e t C(a f o e 2.Ob MPa). oy that the demand for hot water is met and, in addition, it is possible to produce steam with a pressure up to 0.3 MPa for industrial consumers. 38 -Cxj-j for /h« nerf 4nMpt I——I Greif Figur . e HeaSchem13 eth t f o eSuppl y Towe Systeth n f o Greifswalm d from the NPP "Bruno Leuschner" To ensure heat supply also in case of failure of tapping operation possibilite therth s i e f reducinyo e fresgth hf partiao stea d man l operation with reduced fresh steam even at partial load of the turbine. One turbine per reactor is connected with one feed-heating train with three heat exchanger thao s t after completio e firsth tf no construction stage the total thermal power of the four operational units of the NPP . wilMW 5 7 l x amoun A o t t The next step will be to include likewise the units 5 to 8 being now under constructio heae th t n i suppln y system. Thi necessars si o t y have available the load power required for the heat storage units to avoi decreasa d electricae th n i e l powe maximun i r m power-demand times achieve the highest possible degree of security in heat supply, and tap reserve extendinr fo s g heat suppl t presenA y . beyontMW 0 30 d relevant investigations are conducted. The heat is transported from the nuclear power plant to the town Greifswald throug htransia t pipelina e d havinan m m diametea g 0 80 f ro lengt Greifswaln I morf ho e. thakm 0 dn2 therstatioa s i e n where eth heat is transferred to the network of the town. As said, the returning temperatura w no wate s ha rf 80°C eo , afterward wilt i susee r lb fo d heating a planned greenhouse, it will decrease to 60°C. For recirculating the hot water in winter time there are special pumps whose spee s regulatei d e recirculatioth y b d n frequencyr Fo . operation in summer there are again pumps whose speed is controlled by throttles. The maximum recirculation quantity of water is 2 600 t/h. 39 e pipelinTh filles i e d with deonized water fronucleae th m r power plant minimizo T . e corrosion degased deionized wate mixes i r d with soda lime. The resulting pH value is 8.5 to 9.5. The total volume of the pipeline amounts to 28 500 or*. (c) Assurance of radiation protection and nuclear safety Nuclear safet d radiatioyan n protection e ensurehavb o t en i d all cases of heat supply system operation. In conformity with the Regulations of the ODK a documentation concerning the heat supply from the Griefswald nuclear power plant was submitted to the National Board for Atomic Safet d Radiatioyan n protectio approvalr fo n containt I . n sa incident analysis and the description of specific safety and monitoring systems. In the following the most important technical measures of radiation protection are described: The maximum pressure in the hot water system is 2.9 MPa, the minimum pressure on the intake side of the recirculating pumps is 0.7 MPa. À special pressure contro watet ho l re syste th syste n i m m always ensures an average pressure of at least 1.5 MPa which is thus always higher than the maximum pressure of the extraction steam of the third saturatee stagth f eo d steam turbine (i.e 1 MPas i .1. )t I . impossible that the heating steam enters the hot water system because its pressure is always lower than that of the system. If the watet ho pressur e r th syste f eo m approache extractioe th o tha t sf o t n steam, the heat exchangers are automatically put out of operation. e activit extractioe Th th watet f o yho e rwatene th th steaf ro d man system is continuously controlled. If the values reach the permissible limits, the corresponding heat exchangers are automatically switched off. The hot water activity was limited to 25 Bq/1. e tightnesTh heaf so t exchanger s continuousli s y monitorey b d measurin electricae th g l conductivit condensee th f yo d extraction steam. As the pressure of the hot water system is higher than that extractioe oth f valuH p s ne it stearange d , 5 an m. y s betweed an 5 b. n leakage heae th t t exchangersa e detectablesar . e stea degasinTh r fo m feee th gd wate produces ri separata n i d e system consisting of a preheater and a flashing vessel. by that the theoretical possibility of transferring radioactivity to the hot water system is excluded. - The heat transfer station in Greifswald and the activity control in networ e towe th th n f constitutko additionan ea l safety factore Th . incident analysis showed thaheae th tt supply system cannot causy ean disturbances which woul beyono g d d those e th basie taker th sfo s n a safe operatio nucleae th f no r power plant. ) Reductio(d electricaf no l power loss Avoiding electrical power los maximun i s m power-demand times i s a further essential aspect in designing the total heating system. By admixing the hot water with the returning water the return pipe can be use s heada t storage temperature Th . e returninth f eo g watere b wil t lno higher than the maximum permissible temperature of the hot water on the intak recirculatine sidth f eo g pumps, i.e. 130°C. 40 The operatio f fouo n r cross-over station s i optimizes a compute y b d r n dependenci e duratio th e maximun th o e f no m power-demand times e heath , t demand and the travel time of the hot water from these stations to the nuclear power plant. The projec f heao t t supply provide a systesx tank si f mso witha volume of 500 nr* each to store the hot water in case of start-up and shut-down operations and of temperature changes, if necessary the tank water is fed into the system or recirculated. The tanks are designed as low-pressure storages of water having a maximum temperature of 130°C They are filled out of the maximum power-demand time and emptied during this time. (e) First operating experience In the period 1983/84 the heat supply of the town Greifswals reached a thermal power of 50 MW. The ystem has proved its functioning and met the requirements of nuclear safety. Experience gained will be used for further extending the heat supply system. It was stated that the electric power loss without storag s i e0.1 7 r MMW Wpe f therma o l powen i r p operatiota cas f o e 0.3d nan 2 M Wn casi f operatioo e n with reduced fresh steam. These data wil use e r b subsequenl fo d t optimization processesn I . the nuclear power plant and in the district of Greifswald the necessary measures were taken to enable a heat supply with a thermal power of 150 MW durin e heatinth g g period 1984/85. 4.7. GERMANY, FEDERAL REPUBLIF CO 4.7.1 Status of Nuclear Low Temperature Heat Application in the Federal Republi f Germano c y More than 80% of the current district heat capacity of about 33 GJ/s installed in the FRG is being served by fossil fired heating or co-generation plants. Accordin o theit g r parr heafo t t production coal ranges in front of gas and oil. Currently nuclear energy plays no role as an energy source for district heating. l pricoi Aa e resulse th increasef o t e 7o'th sn i s more intensive investigations were carried out as to the possibility of using large nuclear power stations als r generatioofo f heano r distric fo t t heating. With this type of heat generation the primary energy can be utilized more effectively by using the combined heat and power co-generation on technology (CHP). f centralizeI d heat generation fro mnucleaa r plant replace a largs e number of separate boilers or a central fossil-fuelled boiler this also reduces the pollution associated with the work of individual boilers. This advantag usualln ca e y e onl th achievee yb n i e NP th s i P f i d vicinity of a relatively large district heating network, however this is not the case for most of the sites of large KPPs in the FRG up to now. In orde o circumvent r e probleth t f constructinmo g large plants near local concentrations reactor f smalleo s r size se use r b whic fo dn ca h combined heat or power supply or for heat supply only have been developed by the reactor manufacturers. 41 4.7.2 Heat suppl watey b y r reactors (a) Cogeneration With respect to the temperature level required.district heating systems can be supplied by heat from all types of water reactors. Since water reactor sw unde whicno e r e har constructio th f o e fr'Ke th ar G n ni PWR-type the possibility of cogeneration with this reactor type is described here: For the 1300 MWe KWU standard PWR,which are now constructed and licensed according to the convoy principle, a maximum heat extraction of about 500 MJ/s is possible. This maximum heat output is limited by considerations concerning openings in the turbine casing, steam velocity and allowabl ew pressur stresselo e th en i sturbinese blade th f o s . extractee b Hean ca t d froturbine mth thret ea e pressure stages correspondin temperatureo t g abouf so t 18U°C, 140° 115°Cd Can . According to these temperatures the maximum heat extracting is 240 MJ/s, 130 MJ/s and 120 MJ/s. The amount of heat extraction in this range can individually be designed with respect to local circumstances. (b) Heating reactors Currently there are two concepts of light water reactors studied or under developmen FRUe th ,n heatinU i t i.eKW e o s .gth e reactoth d an r called hKRK concept: — — Secondary circuit Heat exchangers -Chimneys -Spent fuel rack -Pressure vessel - Containment Biological shield -Room for fuel element Inspection Figur . Primar14 e y SysteP mNH Arrangement MW 0 20 f o t 42 /!/ The concept which is under development at KWÜ since 1980 is an integrated natural circulated water reactor on the BWR-basis with a 0 GJ/a powebuile 10 b o r st f n ro tfo ca t i primar d an ya MP pressur 5 1. f o e 0 GJ/s e mai50 Th .n characteristic Ktoe th U f heatinso g reactor (see Figure are) 14 : n intermediata - e loop betwee leakine th n g nucleae networth d rkan heat sourc orden i e proteco t r t consumers - the reactor core and the primary heat exchangers are accommodated in the same vessel, which is surrounded closely by the containment; thucore th se will always remain covered with watere , th eve n i n event of loss of the primary coolant - the natural circulation in the primary system and in the residual heat removal loops will function reliably even in the event of accidents o powe,n r suppl. needes it i y r fo d lone th g residua - firse l th tim tf eo assemblie s considerably reduces the space tor storage, allows the raks to be stored within the reactor pressure vessel which results in a compact reactor building. The development costs for this reactor concept including tests for a specially developed hydraulic contro drivd ro l e wil finishee lb 1987n i d . The safety assessment of this concept has been performed by "Gesellschaft fuer ReaktorSicherheit" which results in the statement that the concept is expected to meet the requirements of a German licensing process. e otheTh r / concep/2 t h£Rh whic s studiei h Gesamthochschuly b d e Essen, is an integrated natural circulated reactor on PWR basis. It is designed for a power of 300 GJ/s and a primary system pressure of 1.5 MVa. 4.7.3. Heat suppl higy b y h temperature reactors High temperature reactors (HTk) can be used in different ways in chemical processes and in the production of process heat and electricity. In the FKG the BRC/HRB group and the KWU Interatom group offer hïk concepts which can be used for cogeneration for district heating /!/ HT R designed by the BBC/HRB group nucleae Baseth , n o dMW r ThTd 0 powean 30 R W rM 5 plant1 R ÀV s both designed, constructed and commissioned by BBC/HRB, the BBC/HRB company group conceived advanced Pebbl d higeBe h temperature reactore th s HTR-100, the HTk-300 and the HTR-500, all of them suitable for combined electricity and process steam or/and district heat generation. The Pebble Bed HTR is an advanced nuclear reactor of a very simple design (see Figure 15). The nuclear heat source consists of a loose bed of tennis ball size fuel elements formin reactoe th g r coree sphericaTh . l fuel elements of 6 cm diameter consists of graphite in which the fuel is embedded in the form of coated particles of less than 1 mm diameter. The purocarbo d siliconan n carbide coating fuee th l n particleo s e providesar d for the retention of fission products. 43 Figure 15. Reactor Pressure Vessel with Internals The high inheren te simplsafetth d ean y handling rende e Pebblth r e Bed HTR ideally suite r installatiofo d n e consumeclosth o t e r areas. The plante constructel b sizeal n f o sca s operated dan s twia d n systems. The BbC/HRß company group can therefore offer a complete range of standardized rilk plants from 100 to 120U MW, all suitable for co-generatio f electrio n c power, process steam and/or district heat, 'ihe common featur l e integrateplantal th t o s ei s d design, e.g l component.al s of the primary system are contained in the reactor pressure vessel. Approximately one third of the generator power can be extracted as district heat requiring only minor structural adaptation s comparea s o t d e electricitth y generating plant. Bleedin e steath mf go turbinea s i s suitable procedure for heating water for a district heat system. Main data for electricity generation and co-generation of electricity and heat for HTK-100, HTK-20U and HTRr300 are introduced in Table b. For evaluating the economics the mid-1983 price level of district heat generation from an tiTR compared to that from a hard-coal power plant, e respectivth e electricity generating costs were takebasia r s na fo s evaluating the make-up electricity. This assumption is permissible in so far as in both cases practically the same data of the steam power process can be taken as a basis. 44 TABLE 6. MAIN DATA FOR ELECTRICITY GENERATION AND COGENERATION OF ELECTRICITY AND DISTRICT HEAT Nuclear Power Plant HTR-100 HTR-300 HTR-500 Thermal Reactor Power, MJ/s 256 758 1264 Electricity Generation Electric Power Net Output, MW 100 300 500 Net Thermal Efficiency, % 39 39,6 39,6 (regeneratio f Electricitno d an y District Heat Supply Temperature — — 0 14 - —0 —8 Electric Powe Outputt rNe W ,M 5 9 97- 6 28 29 3- 7 47 48 8- District Heat Output, MJ/s 35 110 180 Overall Plant Efficiency, 1 51,6 - 50,8 53, 52,- 2 2 52,9 - 52,0 (Electricity * District Heat) The district heat generation costs, which vary depending on the supply temperature e summarizeb n ca , s followsa d : District beat Generating Costs: (DM/CJ) HTR-100 7, 6- 11, 5 HYR-300 4,0- 7,0 Hi'K-500 0 5, 3, 0- Hard Coal Fired Plants: (.ßasis: Domestic Hard Coal 250 Uh/t) 100 MW 7,u - 300 MW 4,3 - 7,4 625 MW 3,8 - 6,6 e seeb n ca froe results t mth I , that HTKs wit a hpowe r outpuf o t 300 MWeqU> show clear economic advantages as compared to conventional hard coal fired e conditionplantth r e FRtifo th s .n i s 1 HT12 R Module concept designe e K-WU/Interatoth y b d m group The HTR concep s i baset n KWU'o d s experience wit s wateit h r reactors as well as on the construction and operating experience ecqui red e Arbeitsgemeinschafbth y t Versuchsreaktor Gmbh (AVR) wit e AVR-reactohth r in Juelich, FRC. Special design features lead to a power rating of 200 MJ/s per module for the steam generating version with 700°C core outlet temperature cross-sectioA . modula. a f no r uni f thio t s HT Rs i shown i n figure 16. 45 Figur . Cross-sectio16 e modulaa f no r unistear fo t m generation Main design features of the k-WU/Interatom hTk are as follows.1 standardized well-proven graphite balls wit w enrichelo h d uraniue mor used as fuel element (.pebble bed core) by selectin relativela g y small cor ereactoe diameteth , n 3m rca f ro be controlle y absorbeb d r rod d smalsan l absorber balls e movinth n i g graphite reflector. This permit e "gravitth s y insertion rode th s f "o t pebblho int witd e eo neebe metallir oth hn fo d c cladded incore-absorber rods driven by force. o smalt e l Du core diamete combination i r n corw witlo e ha power density of 30 MW/m , the maximum fuel temperature does not exceed 1600 C during all accident conditions. This in turn precludes a significant failur fuef eo l element particle higa d h an sreleas f eo fission products from the fuel. In the event of a failure of the main hea possibls i t t unii dissipato , tt e decae th e y heat froe th m reactor core in a passive way to multiple redundant surface coolers located outside of the reactor pressure vessel via heat conduction and radiation. Active, redundant primary cooling loops, with associated redundant secondary cooling lines and emergency power supplies can therefore be omitted. 46 applicatioe Foth r UTR-Module th f no e concept with steam generators for district heating several advantages can be claimed. The flexibility choice ith n f poweeo r excellenratinge th d san t safety characteristics allows urban utilities to build and operate their own nuclear power plants for district heating and electrical power supply - even in the direct proximity of towns, furthermore the plant size can be selected to meet the numbee demanth moduley f rb do s also afte demane th r d increases. 4.7.4 Projects carried out in the i?Ku Heat extraction fromulti-purpose th m e natural uranium fuelled, heavy water pressurized water research reactor (M^k; heatinr ,fo e tn g Nuclear Research Centr Karlsruhet ea , started operatio I96n i nd 0an provided the Karlsruhe NKC with heat from 1978 to 1984. The reactor which is out of operation since 1984 had a power rate of 200 MJ/s and provided MJ/s0 2 grie districe th f d o hea.Th r fo t t heating syste operates i m t da a return temperature of 80°C and a supply temperature varying between 11U°C and 13u°C according to the outside temperature. With regard to power production, the reactor was operated as a base load plant. Over the lasoperatio s year5 it 1 tachieves f ha so t i n vera d y good average availability of about 80% also during the time when being operated as a co-generation plant. 4.7.5. Heat extractio process a ne frous StadP sr NP mhea fo e t P StadNP e wit ha therma l outpu 189f o tt 2ne MJ/a d san electrical output of 630 MW has been supplied electricity since 1972. Since December 1983 it has supplied heat for a salt refinery which is locate m fro nucleae k distanca th m5 t a d1. rf eo powe r plant heae Th t. requirement of the saltworks is about 210 000 MWh/year. The steam supply froStadP NP mdesignes i e0 t/h 6 f whicr o , fo d h howeves i rh t/ onl 5 4 y necessary for the saltworks. The remaining 15 t/h is used for space heatinSchilline th t a g g oil-fired power station nearby, during load n immediatela r periodfo d an sy adjacent tank storage facilitye th n O . steae basith mf so conditions , 19u° d 1.0represenh Can t/ 5 0 6 hi"a e ta ,th powe abouf o r 0 MJ/MJ/0 4 t3 r sso respectivel sale th t r refineryfo y . This powe saltworkre th use y b d extendes si 700o t d 0 h/year. Sinc installatioe th e heaf no t extractio f 1983o d n, en statio e th t na the power station had achieved an exceptionally high time availability of more than 80%. 8 4. HUNGARY The yearly fuel consumption for district heating purposes is at presen y hydrocarbons b t , coverearoun% PJ 67 0 t 13 da , which represents roughly 10% of the total energy demand of the country. The number of remotely heate 000d0 e heateflat56 th , s i sd spac f publio e c institutions t a abou . IQ*s i 20 *t m-*. These need e coverear s heay b d t supplied from 344 separate heating systems, 192 of them being of the capacity less than 10 MWt, other 73 of less than 30 MW/t. North Pest is the only district, where the heat demands are of the order of 400 MW/t and Debrecen is the town, where 100 - 150 MWt heat needs are concentrated. These demands could be covered with nuclear heat, considering that a maximum economic level of nuclear heat supply is around 30 - 35 % of tie peak load of the heating system, and the rest of demand should be covered by conventional 47 heating sources, heat demands of further communal heating systems are well unde 0 MW10 r t nominal capacity. Nuclear heas beeha tn utilize e districth n i d t heatin a par f f o o gt e Pakth s city heaÀ . t s outpubeeha nf J 0,12 o P tsupplie 1 o livint d g quarters (1553 o publit flats J 0,14d P can 1 ) institution a e year n Th i s . total heat e demancit th s yi 38,f o 1d e peaMWth k t tfoua e loadTh r . units of the nuclear power plant could provide the consumers from 1988 with around 330 KWt heating power, without any limitations to electric power generation. f Thercourso s i furtheeo n e r heat e aredemanth an i d e nucleaoth f r power plant, therefor developmenw ne o n e n thii t s respect cae expectedb n . At present no other convenient concentrations of demands for heat and electric power provision exis n Hungaryi t intentioo thero n s , s i e o t n build nuclear heating plants in the near future in the country. 4.9 SWITZERLAND As the demand for heating power in Switzerland is not expected to grow dramatically in the next years, the main scope of nuclear district heatine substitutioth s i g r fossilnfo ee las th fuelst n yearI e . th s country was confronted with ecological problems like forest death, acid rain, etc. There is therefore a considerable effort towards reduction of fossile fuel consumption. Two different direction e thereforar s e followe e presenth t a d t time in Switzerland: - heat co-generation from existing NPP introductio f smalno l heating reactor e heath t n i markets . 4.9.1 Heat co-generation from existing NPPs With respect to nuclear heat cogeneratlon, the 15th of November 1983 represents a milestone in the development of nuclear district heating in Switzerland: on this day the first commercial district heating grid from a IxPP in Western turope delivered hot water to the Swiss federal Institute for Reactor Research. Two years later, this network (RtiFUNA) is almost complete coverd e an potentiad th sf o abou% l80 t neat demande th n i s concerned region. e greath Du o tt e succes f thio s s first network several other projects, concerning the remaining UPF's in Switzerland are now seriously discussed. Tabl summarize7 e e maith s n characteristic l thesal f o se projects t I mus . notede b t , thae licensinth t e projecteth f go d Kaiseräugst NPF is bound to the condition to deliver some 0.5 GWth to the nearby sited cit f daseyo r districfo i t heating purposes. 48 TABLE 7. HEAT CO-GENERATION FROM NUCLEAR POWER PLANTS IN SWITZERLAND PR036CT PR03CCT N PP THERMAL SUBSTITUTION COST» NAM6 STATUS POU>€R POTCMTIAL CHW) CMW) t oio/j( r RC FON A Op- 0C2NAU 2 "96o 60-70 -(fi'ooo *.-. 3000 -450 So'ooo 2 7.8. S - FOL A Pr. S0S«N FEMBe FV. MÜHLEB6R6 960 ^,-000 «o-oco TRANSIOAAL Pr. B6T2NAU . .9 M 960 -290 «..-*r (+ J.6IBSTAPT) (BASEL)* Pr.* JCAISERAUÄST* 3000 -4W 350' 000 ___ KAISERAUGSP undet NP ye * rt constructionno s i T . Connection to the heating grid is foreseen by governmental decision. A.9.2 Introductio smalf no l heating reactor heae th t n markei s t (a) The Swiss Federal Institute for Reactor Research at Wuerenlingen manr fo (FIRy s year;ha s followe developmene th d nucleae th f to r heat market. In 1983 a study concerning a small heating reactor (originally still with homogeneous fuel beguns )wa orden ,i provido t r ea loca l heating network (community or region) with the produced heat. R concepfI gooe d Th tha d chance f suco fulfilo m systemha t s ai e lth , i.e. high substitution potential w burdenin lo ,environmente th f o g , high safety and competitivity against conventional heating systems. In 1984 the ideas about the reactor concept became more concrete and in autumn 198 4a firs tdesige drafsmala th f f no o tl heating reacto s beerha n presented. e ide Th nucleaf ao r district heating with small heating reactors sha generated much interest, i'he continuation of the FlR-study "bwiss heating Reactor (SUR)" has been approved and many representatives of private industry have decided to participate actively in the deeper examination of the concept e higTh .h public interes Silk-Concepe th n i t t motivated then on the one hand the Swiss Institute for Nuclear Research (Slh) to develop an alternative heating reactor concep tothee (GEYSERth rn o han C d BB d)an and the daughter company HRb to propose the study of a gas-cooled, pebble-bed heating reactor (GHR). The technology of the SUR is based essentially on the well known LWR-technology, that of the GHR on the HTR-technology. Both are well tested technologies and there is much experience with both of them in Switzerland. The concept of GEYSER on the other hand is very similar to Swedish and American experimental reactors used for basic research. 49 The private industry is represented for the two projects SUR and GHR by Bonnard & Gard el, BBC Brown B over! & Cie. , Elektrowatt, Sulzer Brothers and Motor Columbus. Togethe theR r FI wit ye decidehth 198n i d staro 5t a t common research programme for the development of a small. heating reactor. The aim of this programme is to design in a first phase a heating reactor syste examinn detaio i mt d lan e this design froeconomie th m d can licensibility poin viewf to . Following thi decisiosa n abou detailee tth d project work, sponsorin d credian g t applications wil madee lb orden ,i o t r construct a demonstration plant on the FIR-site (.or elsewhere.). The related contract, regulating also the confidential aspects and the knowledge l transfery 190al Ha y 5f b o bees d ,ha en n e signeth t a d participants mentioned above. The convergence of all three projects (SHR, Gdk and GfcYiUR) towards the presentatio preliminara f no y repor f similato r degre developmenf eo t f Julo yd 198en e assureas 6th ti a commo y b d n coordination committee. (b) Characteristics of SHR, GHR and GKYSER e result e firss baseth i th n R to f d o sSU e concepe Th th f o t study, presented in September 1984. The heat produced in the reactor core is transferre naturay b d l circulatio primare th f o ny water througo htw heat exchangers (integrated in the reactor pressure vessel; to two intermediate circuits o secondarTw . y heat exchangers enabl heae o eth t be transferre districe th o t d t heating grid e reactoTh . r pressure vessel (core outlet température 198°C, pressur bar5 1 e undes i ) r watelarga n i re water pool, whic s tightlhi y close concreta y b d e containmente Th . reactivity control occurs by poisoned fuel pins (burnup compensation) and control rods wit hhydraulia c driving mechanisme th f e o fue Th s .li LWk-type, with low enriched primare th t p ye coolin ne GHRe o c th - n gI mediu Heliums i m , whics hi forced by a circulation pump to pass through a graphite moderated pebble-be undeC 0 pressurra heates d45 i cor5 o d 1 t dean C f o e fro O 25 m bar. The heat is then transferred to the district heating network by means of an intermediate cooling circuit. The reactivity control is achieved by means of control rods, inserted into the graphite reflector. The fuel consists of the same spherical fuel elements which are foreseen for the German UTR reactor. GEYSIe Ideem th n 0 p5 Rbottoe wela concepth f t lo ma core s th ti e which contains highly borated water. This well forms also the reactor containment, made by reinforced concrete. The depth of the well results from a hydrostatic pressure of about 5 bar in the core region and allows therefor core th ee outlet temperaturee reactoTh . C raboue b 0 o 15 t s pressure vesse alss li o essentially replace watee th ry b dcolum n existing above the core, ïhe heat generated in the core is transferred through a diffuser to the steam condensers, ïhe intermediate circuits work on the principl thermosyphone s th a f r o e fa s desige (a Th . e nus doe t sno possible; ay active components for the reactivity control as well as for the main and decay heat transfer. Inherent changes of the boron concentration of the "primary water" flowing through the core is used for reactivity control e corTh e. desig stils i n l under discussione b n ca t I . an SHR-similar TRIGA-similaa cor r eo r concept. 50 4.10 USSR 4.10.1 Introductory remarks Calculations performed by specialized institutes allow to expect that onle increasth y n heai e t consumption would amoun abouo t t t 2.10' Ccal by the year 2000, which implies that if each power unit provides generation of heat power equivalent to an installed power of 1000 MW, at least 60Ü power sources would have to be constructed. This problem cannot be solved by using power source of only one type; therefore for the nearest future the following possible power source e considerear s r districfo d t heating: - fossil-fuelled heat sources (.gas, solid fuels,oil; nuclear powered heat sources (a) based on co-generation of electricity and heat (ChP; from NPPs: 1. using uncontrolled steam extraction 2. using the TK-450-500/60 type of turbines (developed for co-generation) (b) based on specialized nuclear heating plants (NHP) of the type AST 4.10.2 Status of nuclear heat co-generation in the USSR At present the heat gained at NPPs by the co-generation technology (CttP; is supplied to nearby consumers. The heat in a form of hot water feed e districth s t heating system coverd an s s also, warm water consumption in the NPP personnel settlements, adjacent population centres, factories etc. In the USSR the co-generation technology is used at several NPPs in different sizes and on the PWR-type as well as the RBMK-type of reactors (see for example the IAEA TECDOC-397 "Potential of Low-Temperature Nuclear beat Application", lAhA 1986). Tabl show8 e n overviea se heath f tw o suppl y fro me Sovie th som f o et e yearth NPP n si s 1981-82. TABL . 8 E HEAT SUPPL VARIOUY B Y S NPPs (1981-1982) NPP Thermal capacity Heat consumption Gcal/hr Gcal/yr Beloyarks 40.0 0 9540 Kursk 265.0 7 20 5 46 Novoh z - ne V o r 152.0 1080 000 Kovno 23.8 124 886 Armyanskaya 26.6 4 1411 9 Kolskaya 39.2 343 480 SI The type of condensation turbines currently used at Soviet NPPs, designe r differenfo d t amount f heao s t extraction s introducei , n Tabli d . 9 e TABLE 9. PLANNED HEAT SUPPLY FROM OPERATING TURBINES Type of turbine Number of turbines Thermal supply per 1 reactor capacity Gcal/hr K-220-44 2 5 2 x 2 K.-500 -6 5/3000 2 2 x 50 0 K50 -1 1000/ 6 0- 1 K- 7 50-6 5/300 2 2 x 100 The nuclear power plants, already foreseen and designed for electricity and heat co-generation (CUP;, can provide significantly higher amount f specia o f heate o s Us .l turbine t suca s h CriP-NPPs permite th s heat supply from one 1000 MW(e) power unit to be increased up to 900 Gcal/hr e firsïh . t i\P f sucPo a htyp e wit n electricaa h l poweG2 W f s i o r planned for district heating of the city of Odessa; this plant will have two units equipped with one WWK-1000 reactor and two Tk-450-500/60 turbines e totaTh l. therma e citl th loay f o dcovere nuclear-producey b d d heat in combination with fossil-fuelled peak boiler units will amount to about 3000 Gcal/hr in the designed coldest regime. The ChP-NPP uses the same WWk.K-1000 reactors which are installed at NPPs, therefore the limitations for siting of these plants with respect to the location of large industrial and population agglomerations are similar LO those for l The Odessa ChP-NPj froe s citym i sitek fth m 5 d2 , what requires thaa t great heat feeder be build (about 60 000 t of 1000 mm diameter pipes for heat transit from the plant to the peak boiler units in the city, where the nuclear-produced heat enter e heath s t distribution system. e yeaBth y r 199e constructioth 0 f similao n r CHP-NPP s i envisages d also in the cities of Minsk, Volgograd and Kharkov. 4.10.3 Utilization of the Soviet NHP of the AST-type Several specific conditions, suc s deficia h f servico t e water, ecological reasons a simultaneou neeo r n ,fo d s suppl f largo y e electric and thermal power, some economical reasons and construction schedule, make t necessarno t i o built y d large CHP-NPPs r thesFo . e MW(t 0 case50 )a s water-water vessel-type reactor (K4 AbT-500; has been developed, designed for generation of heat in the form of hot water. AST produces 860 Gcal/hr heat as hot water and, in co-operation with peak-heat sources can provide heat to a district having a total heat consumption of the order of 1700 Gcal/hr. 52 e verth y unf o eimportan t prioritie f thio s s l*i s i Pthat , with respect to the inherent safety principles used in its design, these station e buila closb n n i de ca s vicinit e populateth f yo d agglomerations. The requirement r constructiofo s sitind nan f thesgo e th NtiP n i s USSR were issued alread e n "Genera197annen i yth a s 8o t a x l Regulations for the Safety of Nuclear Power Plants During Design, Construction and Operation" Englise Th . h translatio f thino s document s i introduce, n i d this repor s Annea t . 2 x The "Requirements" take into account several important aspects, such as: measures excluding core melting during damages resultin n Li g OCA - external impacts connected with human activities, crashed airplanes, shock waves, etc. technical precautions related to fuel element reliability, spent fuel transport, heat transfer, reactor circuits' temperature, heat exchanger location etc. 4.10.4 Description of the AST-500 NHP*) Aa resuls f investigatioo t n studies which reactor type coule b d a reliable a basiNH r Pfo smeetin l safetal g y requirements introducen i d the "USSR Siting Requirements .." (see Annex 2) a vessel-type water-water reacto e bes takes th ts wa ra n candidate structurae Th . l schem f thio e s reactor called AST-SOU s i showmaie , Th n Figuri n technica . 17 e l datf o a the ASÏ-500 NriP are introduced in Table 10. «Note; Accordin e participant th e wis th f o h o t g f thialso d se previouan sth o s IAEA meetings dealing with the nuclear heat application, the description e Sovieoth f t AST-500 NH s i Pintroduce n detaili d s becausa et therno s i e comprehensive published English text availabl thin o e s subject. 53 Figure 17. Reactor unit: 1) reactor; 2) cor«; ) insuranc3 e vessel shel) A ; l uni f pipeo t s and equipment; 5) pipeline» of secondary circuit; 6) pressure compensator; 7) con- trod safet an ldrivd ro y e Mechanisms- ra ) ;8 diation shield ) genera9 ; l contro d secro l - tion ) heat-exchangers10 ; individua) 11 ; l contro d tubes ro lIntravcsse) 12 ; l pit) ;13 rotation device. Tabl. 10 e MAIN TECHNICA E LAST-50 TH DAT F AO 0 REACTOR PLANT Characteristics Dimension Numerical value Reactor thermal power MW 500 Characteristic primare th f so y zil'C- circuit coolant: inlet pressure/temperature MPa/°C 2.0/250 (130) outlet under-heating to saturation 0/10.0* « outlet void fraction 0 2.0o ut p0. / Characteristics of the intermediate circuit: pressure MPa 1.2 " l-II circuit heat exchanger inlet/outlet temperature 90/170 Characteristics of the heating grid system: pressure-, heat exchanger MPa 2.0 a MP heating grid 1.6 supply header/return header temperature 150/70 (140/60)* Characteristics of the core: specific power density MW/n 27.0 fuel element diameter/typm m f fueo e l 13.6/UO, reactoe th - Datr r* fo a operation under unboillng conditions. 54 (a) Construction of the reactor facility: The reactor unit (Fig.17), which is the principal part of the reactor facility (RF), represent assembln a s e constitutiof planth o y n i t f no whic s includei h water-cooled/water-moderatea d d integral-type reactor e purposwhichth r f localizinfo o e, e hazardth g s associated with depressurization of the reactor vessel or the pipelines of the ancillary system of the primary circuit, is enclosed in an "insurance" vessel. Th e heat-exchangere th cord an e e primarth f d secondaro s an y y circuits are located in the reactor vessel. The upper space of the reactor above coolane leveth th ef o l t perform e functioe th s th f o n primary circuit pressure compensator, heat removal froe cormth s i e effected by natural circulation of the coolant. The ascending part of the circulation circuit include a sectios n wit e individuath h l controd ro l tubes and, located above them, the general control rod section. The general contro sectiod ro a shafl s i nt with little coverin structuray b g l components. À uniform supply of coolant to the tubular systems of the heat-exchanger s providei s d through opening e pipe sshelth th made f o n li e and equipment unit. The heat-exchanger e primarth f secondard o s an y y circuite ar s arranged uniformly in the gap formed by the intravessel pit, the pipe and equipment unit and t fie reactor vessel. The heat-exchangers of the secondary circuit pipelines are incorporated in three loops, which in the event of leakage of the tubular systems being detected, can be out off by inle outled an t t gate valves, located directl e insurancth n yo e vessel. The reactor within the insurance vessel is mounted on an annular bearing e insidth n i e, recesse f whico s s arrangei h e radiatioth d n shield. In the intervessel space below the annular bearing, the rotation devic s i locatede , designe e assembl th e suspende r th fo df o y d equipment for the remote control monitoring of the metal of the reactor vessel and e insurancth e vessel. Abov e annulath e r bearin e locatear g e pipelineth d e secondarth f o s y circuit, the pipelines of the ancillary systems of the primary circuit, the cables of the primary measurement transducers (suspensions of ionization chambers, level meters, heat transducers) uppee Th r. flangf o e the radiation shield, which beyon e insurancth d e vesse s i coverel d with the iron-concrete structures of the reactor pit, is located above the reactor cap. Abov e uppeth e r e radiatioflangth f o e n shield, insid e insurancth e e vessel e locatear , e controth d safetd drivd an l ro y e mechanismse th , e electricableth f o s c motor positiod an s n sensor e controth f d o s an l safet actuatorsd ro y als d e e cableintrareactoan oth ,th f o s r measurement probes. The reactor vessel is a hermetically welded container with two flanged joints, which consists of the lower and upper parts of the vessel and the cap. A hollow block is designed for the location in it of the fuel elements of the core, the control rod tubes, the mechanisms connectin e absorbinth g g rods wit e controhth safetd an d actuatorsl ro y , and the guides for the intrareactor monitoring equipments. 55 A straight-tubular heat-exchange e AST-50s choseth wa r fo n0 reactor, with movemen e coolane primarth th f o f t o t y circuie intertubth n i t e space. This type of heat-exchanger allows the heat-transfer surface to be more efficiently planed in the reactor space, bounded by the core pit and the vessel structures. In order to organize the collant flow of the primary circuit, each heat-exchanger is provided with an individual casing vere Th y . slight pressure drops betwee e circuitth n s determine th e minimum level of the mechanical stresses in the heat-exchanger structures. (b) Basic circuit e reactoth f o s r facility The primary circuit, intermediate circuit e mainth , s water circuit, intermediate circuit for internal requirements, and the technical water circui e include ar e constitutiot th n i d e reactoth f no r facilite th d an y servicing systems (Fig.18). The primary circuit includes the main circulation circuit, and also the systems for the coolant pressure compensation, coolant purification and purification of the feed water to the control and safety rod mechanisms in order to prevent the steam-gas mixture entering them. This circui s i providet d with system r fillinfo s g and makeup, sampling, air removal, and drainage. Fig.18. Schematic diagram of the reactor facility: 1) primary cir- cuit makeup system ) secondar2 ; y circuit shielding system ) pres3 ; - sure compensation system; 4) mains water system; 5) secondary circuit ancillary systems ) intermediat6 ; e circuit primar) ;7 y coolant cir- cuit ) syste8 ; purificatior fo m f coolano n d feean t d e wateth o t r control and safety rod mechanisms. e secondarTh y circuit includes three autonomous circulation loops. Its structure includes systems for the circuit pressure compensation, purificatio coolante th f no , fillin d makeupan g r remova,ai d blow-offan l , protection of the circuit from excess pressure above permissible. e maiTh n water circuit withi statioe th n n bounds includes three circulation loops (correspondin three th e o g t secondare loopth f so y circuit), joined colan d t dwate ho wit e rth h manifolds e technologicaTh . l e maintainear parameter F R e e supplmeany th th b d f mainf f o yso o s water 56 through the network heat-exchangers by the installation of a regulating valve. The normal cooling of the reactor facility is effected by circulation f technicao s l water throug e networth h k heat-exchangere th r o s cooling heat*exchangers. ) (c Core The core is assembled from 121 fuel-element assemblies according ttriangulaa o e fuel-elemenr Th gri . d mm wit 3 pitca ht24 assemblief ho s contain bundle fuef o s l elements wit diameteha f 13. o rcontaine, 6mm n i d hexagonaa l sheat zirconiuf o h m alloy wit siza h e "abov 8 keye 23 eth "f o mm and a thickness of 1.5 mm. Individual gravity tubes are provided in the designwhich ensure hydraulic profilin coolane th f o gt supply through the fuel element assemblies according to their thermal loading. The fuel element is made of tubing with dimensions 13.5 x 0.9 mm of zirconium allo d filleyan d with pellet bakef o s d uranium dioxidee Th . fuel element fuee th l n elemeni s t assemblie arrangee verticee sar th t a d s of a regular triangular lattice with a pitch of 17.8 mm. For the partial compensation of the reactivity margin on burnup and r profilinfo powee th g r distribution ove e coreheighe th th r ,f o t absorbing elements are installed in the fuel-element assemblies in place x fueosi fl elements. a tub Thef zirconiue o e yar m alloy with dimensions of 13.6 x 0.9 mm, filled with lump boron in an aluminum matrix. In order to monitor the neutron-physics and thermal-hydraulic parameters of the core, special measurement probes are provided. For this, in place of one of the fuel elements in the fuel-element assembly, a guiding zirconium tube is installed with dimensions 13.0 x U.9 mm, which serves for locating the probe containing thermal transducers and neutron flux sensors. The core is calculated for operation in partial recharging conditions, with an interval between them of 2 years. Fuel with a uranium enrichmen d 2.0 firse an 1.0f o th %tuses 6 i tn ,1. i d charge e th n i ; stationary regim operationf o e , makeu effectes i p d with fuel wita h uranium 2.0%d enrichmenan calculatee 6 .Th 1. f o t d core characteristics are given in Table 11. In orde controo t r reactoe th l r power a ,regulato s locatei r n eaci d h fuel element assembly centra(excepe th r lfo t one) consistt I . 8 1 f so movable absorber rods, joined with a common crosspiece. The rods are dispose guiden i d zirconiuf so m tubes with separate, dimensionmm 1 x d8 1 s bundle ith n e together wit e fueth hl elements. Ther e alsear o guides inside the gravity tubes. The absorbing rod is a tube of stainless steel with dimension fille, smm 12.2 d 1. wit 5x h lump f boroo s n carbide. The movable absorbing rods of three or four fuel-element assemblies are joined and form the working element of the control and safety system e reactooth f r (CSSj e systeTh . m consist workin6 3 f o s g elements, eacf o h n actuatorow whic s moves it hi y .b d Withou actuatorse th t absorbine th , g rods are fixed in the fuel-element assembly in the very lowest position by special arrester devices, located there. The CSS possesses sufficient reactivity for fulfilling all the planned operating regimes capabli d f transferrinan so e e reactoth g r from 57 Tabl . Calculate11 e d Core Characteristics Diameter of core, m 2.8 Ü 3. Heigh f coreo t m , Energ0 3 y intensit f coreyo , kW/liter Uranium charge, tons 50.0 Operating time of fuel of first charge, eff. days 460 Power release nonuniformity coefficient: over the volume of the core 2.2 ove e radiuth r s 1.25 8 1. oveheighe th r t ove e fueth r l element assemblies 1.24 Effects of reactivity, keff; by heatine wateth f ro g 0.012 by heatin e fueth lf go 0.011 by 135Xe poisoning 0.023 Margi n burnupo n , compensate controy b d d safetan l y units at nominal power, keff 0.021 Margin on burnup, compensated by the burning-up absorber, keff 0.044 Reactivity margin of fuel of cold fresh charge, keff 0.067 Efficiency of control and safety rod system, keff 0.20 Duration of fuel cycle (with three rechargings), years 6 Depth of burnup in stationary operating regime, 4 1 MW.days/kg any e subcriticastatth o t e le conditiostatth n i e f malfunctiono e th f o n most efficient elemen f e actioreactivityo tth n no . addition I e electromechanicath o t n l CjjS insurancn a , e boron system of e reactivitactioth n no s providedi y s calculatei t I . a n o d hypothetical accident with a considerable nember of CSS elements recorded e verth y n i lowest positio wit d e necessitnan th h f emergenco y y cooling. Pulse-current chambers are used for neutron monitoring, installed between the intravessel pit and the reactor vessel. (d) Water-Chemical and Gas Regime of the Primary Circuit A neutral adjusted hydrogen-helium water-chemica s regimga d ean l (WCGR) is used in the primary circuit of the reactor facility. The high e puritreactoth f o y r water valu H witp ea h clos o neutrat e l during operation of the reactor is achieved as a result of continuous purification of part of the coolant and by carrying out its thorough deaeration during startup. Normalizatio e hydrogeth f no n concentration i n the reactor water at the core inlet is accomplished by starting from the condition of suppression of radiolysis of the water and the formation of radiolytic oxygen in the circuit. Adjustmen e WCGth s i achieveRf o t d with hydrogen containee th n i d composition of the working gas of the steam-gas pressure compensator. *'or this, the content of hydrogen in the hydrogen-helium mixture used is assumed to be such that the mixture cannot form an explosion-hazardous 58 concentratio hydrogef no n when mixey proportionsan dn i wit r hai , whics hi ensured by the explosion safety of the overall station systems. Monitorin qualite th f o gy e primarindeth f xo y circuit coolans ti carrie t periodicallou d meany b y f samplso e analysis. Instrument methods of monitoring the individual parameters are also used, with presentation of the data on a modular control panel. Samplesteam-gae th f so s mixtur e takeear n only whecontene th n f o t watee gaseth rn si (with respec hydrogeo t oxygend nan ) exceede sth maximum permissible values. Continuous monitorin s leakagga f o ge froe th m reactor la carried out. (e) Safety Systems and the Protective Systems of the Reactor Facility The reactor scram system (SS) is designed for stopping, slowing down, or limiting the chain reaction in the case of the onset of emergency situations or deviations from the conditions of normal operation. This is achieved by giving the appropriate commands to the control system of the CSS actuators and by the subsequent insertion in the core or by vetoing withdrawal of the working elements. The insertion with maximum velocity workine th l ogal fS (dumpin elementCS e th gf so wit h deenergizine th f go actuator motors) provides the maximum possible level of protection. e scraTh m system operates upo attainmene nth dangerouf to s power settings of the reactor or its doubling time, in the case of overpressurization or depressurization of the primary circuit, earthquakes statione r idlin,o th f o g , disconnectio .thf no e NUb, requiring withdrawal of the KF from operation, and pressure on the scram system pushbutton modulae th t emergencr a so r y control panels. The number, disposition, efficiency, and injection velocity of the operating element scrae th mf so syste e sucar m h tha hazardoun i t s situations the working elements without exception most efficiently provide the necessary (with the condition of unimpairment of the fuel elements; spee f reductioo d e leve th f powe o nf subcriticalit o ld ran e reactorth f o y . Emergency cooling system e reacto(tiCSth f o j r facilit designes i y d for the provision of the safe removal of residual heat release in the case of emergency situations, related with the impossibility of heat removal by the thermal network and the normal cooling systems. The LCb loops are connected to the loops of the secondary circuit or to the loops of the network circuit. e removaTh f residualo l heat release frocore e meany th mth b e f so emergency cooling system is effected by the natural circulation of the coolant through the core and the heat-exchangers of the primary-secondary circuitcooline th d gsan heat-exchangers, with natural circulatioe th f no cooling water from the reserve water tank through the cooling-heat exchangers evaporatioo t d the e ,an ndu n intatmospheree th o e reservTh . e of water in the tanks is chosen from the conditions for providing prolonged remova f residualo l heat release (not less tha days;3 n . Protection primare systeth f mo y circuit from overpressurization ensures protection from excess pressure above the permissible value without discharg activf o e e coolant y mean,b f reliablo s e heat removal from the core into the secondary circuit and subsequent discharge of the heat intatmosphere th o e throug e coolinhth g channels existine th n i g 59 secondary circuit. This method of protection is achieved because of the large storage capacity of the circuits, their rigorous thermal connection provided by the integral grouping of the primary circuit plant, and also the natural circulation through the primary circuit, the extended heat-transfer surface of the primary-secondary circuit heat-exchanger, the natural circulation through the secondary circuit in emergency conditions, and by use of the standby system of emergency cooling functioning on the passive principle, each of the three channels of which is capable of providing the removal of heat from the shut-down reactor and protection of e primarth y circuit from overpressurization. n additionI a syste, f emergencmo y heat remova s i providel e th n i d facility, using standby pulsed safety devices in the pressure compensators of the secondary circuit in conjunction with a reliable make-up. Opening of the pulsed safety devices in the secondary circuit and the discharge of coolant from it in the form of steam ensure the removal of heat from the primary circuit and protection of the reactor from overpressurization. The openin e pulseth f go d safety device e secondarth n i s y circuis i t provided for both by a signal concerning excess of the permissible pressure of the primary circuit and by the direct action of the pressure of the secondary circuit. In the latter case, the pressure in the primary circui s i limitete valu th y eb dacceptabl e establisheth r fo e d tensile strength of the reactor. Thus, the principle of protection of the reactor is accomplished by means of safety devices, but without the discharge of active coolant. This is very important from the point -of view of radiation consequences in the conditions of location of heat supply stations in the immediate vicinit f largo y e cities. (f) Localizing Systems The insurance vesse a passiv s i l e protectiv localizind an e g device, ensuring safety in the case of depressurization of the reactor vessel within the limits of the design value and in the case of ruptures of the pipelines of the primary circuit system located inside the insurance vessel. A pressure monitor is provided in the latter. The e desiginsurancth f o n e vessel consist a lowe upped f o san r r parte th , volume and configuration of which are chosen from the condition of guaranteeing the level of the coolant in the reactor above the core. The dual shut-off armature system in the primary circuit pipelines within the confines of the insurance vessel is provided for limiting the discharg f activo e e coolant fror e reactoensurinth mfo d e levean rth g f o l coolant in the reactor above the core in the case of accidental depressurization of the pipelines or plant of the primary circuit systems outsid e insurancth f o e e vessel. When the armature is closed, the reactor volume is cut-off from the ancillary system of the primary circuit. The shut-off valves used are relate a typ f armaturo o t de e whic s i closeh removiny b d e electrith g c power sypply from the controlling systems; they have pneumatic actuators with remote and automatic control, and are vacuum tight relative to the outside medium. Together wit e automatith h c valves, return valvee ar s used in the shut-off trunk lines of the primary circuit systems. 60 Information about the position of the driving shut-off armature of the primary circuit, located inside the insurance vessel, is fed constantly to the control panel. With the appearance of signals indicating accidental depressurizatio e primarth n i ny circuit systems (for example, pressure, radioactivity, level;, automatic closure th f o e high-speed shut-off e primarvalvth f o ey circuie insurancth n i t e vessel occurs, and the reactor is cut-off. The slide e valvesecondarth f o s y circuie insurancth n i t e vessee ar l used for ensuring localization of active coolant of the primary circuit in the cas f pressurizatioo e n breakdow e primary-secondarth f o n y circuit heat-exchanger. The slide valves are installed at the water inlet and outlet of the primary-secondary circuit heat-exchanger, they are disposed e insurancth e e shieldevessear d an l d froe outsidmth e effectf o s structural members. Electric drives with manually operated duplication are used as actuators in the slide valves. The slide valves are controlled remotely froe mai mth r emergenco n y control desk. Closure slidth ef o evalve s effectei s n sequencei d e th :l firsal f o t slide e colvalvth dn o ebranc t s i closehho thee d e valvth dan nth n o e branch. In order to prevent spurious closure by the slide valve control circuit, an appropriate interlock is provided. Signalling of the extreme positions of the shut-off element in the slide valves also is proposed, with e informatiooutleth f e controo t th t a n l e reactopaneth f o l r facility. Cg) Protection of the Facility from External Lffects Protectio e reactoth f no r facility from external effects wa s planned by starting from the following conditions.* an aircraft crash with km/h0 70 ,o t wit p a ton0 speeu h2 t f a ssubsequeno o d t ap u mas t o s t possible fire e actio th a shoc: f o nk wave wit overpressurn ha e th t a e 0 sec1 a tim. o t shocd f actio o ep an o u 0.0t a k f p n5o fronu MP f o t The layout of the reactor division of the Oor'kii heat supply station provides for the disposition of the reactor and the systems associated with the primary circuit in rugged-compact compartments, calculated on the external effects and also on an internal pressure which may result in the cas f leakago e f coolano e t froe plan mpipelinesth d an t , causing losf o s pressurization. The protectio f systemno t possiblno s whics i o enclos t et i h n i e compartments, because of their branching and large dimensions (plant and pipelines of the secondary circuit loop), is effected with an emergency system and dispersal siting of them in compartments separated by barriers, calculated on the external effects (schematic view: see Figure p IP ( I I I l i î I I I i A l ^XI^rrliTrtyriS^rriTi-J^T^.-fef^TM-mlTrylTT il^||4jrilif«A*f AST-50 o e NHth Tw Figur Pn i Conno 0. 19 e n Building 61 The safety system, located outsid e concretth f o e e shield, incorporates three independent functional channels, each of which completely ensures the operation of the functions of this system. The independent channel e safetth f o ys syste e spacemar d territorially relativ reactoe th o t er pit. Consequently, direc shadod an t w shieldins i g provided from an aircraft crash in any direction and also from a directed explosion shock wave. The plant of the reactor facility is capable of operating in regions subjec o seismit t ce maximu effectse th cas th f o en m I planne. d seismic effects e reactoth , r facilit s i shuy t dowcooledd an n . Automatic shutdown e reactooth f r fro ma seismicit y senso s i providedr . Thus e structurath , layoud an l t solutions adopted have alloweda leve f e safetreactoo l th f yo r facilit ensurede b o t y , thus permittins it g locatio vicinite th n i nf larg yo e cities constructioe Th . n developer fo d e piloth t unit n Gork i sVoronezd an i s contemplatei h e commerciath r fo d l utilization of AST-500 facilities in the European part of the Soviet Union. Another design of the NHP with two AST-500 is shown in Figure 20, where each reactor unit together loopwitS d safethEC an s y systemd an s spent fuel storag e placear e d insid ea containmen dlam 3 . 3 wit f o ta h spherical dome. Systems and installations such as fresh fuel preparation zone, areas for storage and repair of transfer-handling equipment, liquid waste processing system, repair shops, ventilation centre which serve both reactor units e e maiplacear th , n n di buildin g outsid containmente th e . The containmen e reactoth f o t r par s i designet d with e regarth o t d influence from the exterior due to impact of 20 t airplane fragments with 700 km/h velocity and due to blast wave of 0.05 MPa pressure. «0300 T unita containmenT n i AS s o Tw Figur . t 20 earrangement . 4.10.5 Features of the District Heating System Operation Several studies have bee ne implementatio th mad n o e e th f o n AST-50 0e distric th NHP n i s t heating system , e analysibaseth e n th o d f o s weather conditions in the USSR. To show an example an annual schedule of heat consumptio districa y b na cit f yo t situate e Europeath n I d n parf o t e USS th s i introduceR e schedulmaio Th n tw Figurndi s period. eha 21 e s correspondin o wintet g d summeran r ; between thes maio tw en periods another transitional period exists whe e combinatioth n f heao n t sources e adjusted b e winteha th o t sn I r .period , wit e outdooth h r temperatures 62 IGc»l/h) 1700 1600 1SOO — KiHtl ijp«f>o<« • — l —— 1400 • V h 1300 T r«nlilion«l - M— B«e( MflOrt -- •« — Summer period — ———« pcrKxl 1200 i 1 l tnatcHo r wpply 1100 -K \ *1N 1000 s. 4+I1 '*-' MW^ 900 t * ,,vMVM,, l [^S 800 'nï o 700 1 * v \ T . 1 600 L 3SJ 500 js \ TT Lr 1 400 H4t° |-? 1 300 1—o 1 1 200 -jr Mf r////// '///l/l lit) 1 1 100 . a . J!J_ -? fi f 1 IM / 1000 3000 3000 4000 5000 6000 7000 WOO «7«0« 11 « -3720 -5300 : Heat consumption Im rtadmtWm : Heut from AST.un ///////////f Figure 21. Annual Chart of Heat Output from an AST and from Peaking Facilities. e rangi ti nf -30° o e -1°Co C t base ,th e heat loa supplies i d d froe th m basic power source and the rest is delivered by peak sources for both heating and warm water production. Whe outside nth e temperature reaches +8° summee Cth r operation seaso nbasie beginth cd powean s r sourc s supplyini e g onlwatert ho y s A . seen froschedule th m average eth e valu f thieo s load doeexceet - sno 5 1 d nom.Q % .20 As the outside air temperature rises, the direct network water temperature reduces exampln a ; e forsee networa r nfo k witAST-NHe th h s Pa a heat source (see Figur ) show22 e s typical curve changer e fo sth f so supplied and back water temperatures for the conditions of the European part of the USSR. With fossil-fuelled heat sources the direct water temperature decrease brings also savinge fueth l n i consumptions . 63 DIRECT GRID WATER TEMPERATURE (RATING) UO- AST-HOT WATER TEMPERATURE 90- 70. BACK GRID WATER TEMPERATURE (RATING) SO 13- 5 + 0 5 - 0 -1 5 -1 0 -2 -33 5 -" Figur . Gri22 e d Circuit Temperatur a Functio s a e f Outeo nr Temperatur Ai r e When co—generation units are employed (fossil-fuelled, as well as nuclear) then accordin e heath t o t gconsumptio n load decreas e electrith e c power productio supposes i e increased b e workinP b n NH o ca n t d e n i gTh . co-operation with peak fossil-fuelled boilers; the heat generation can be controlled by changing the temperature difference on heat exchangers. 64 5. CONCLUSIONS The growt f nucleao h r power n termi , f installeo s d capacitys i , proceeding worldwide althoug a slowe t ha r pace thaforesees wa n n i n previous nucleaprognoses% 16 e re world'Th sharth . n i e s electricity production is bringing reasonable savings in the consumption of fossil fuels. The substitution of nuclear sources for fossil-fuelled heating station s proceedini s g slowly. However several countries already weno t t the conclusion that the utilization of nuclear heat can not only bring savings in the fuel economy but contribute also substantially to the cleanliness of the environment. The Advisory Group meetine Nucleath n o g r Heat Applicatior fo n District Heating confirmed that there were already good experiences gained with the utilization of low-temperature heat produced by co-generation at NPPs (CHP). The technology-scheme is fairly simple and the CUP can be taken as a proven technology for both hot water and low-tempe rat u re steam supply econome Th .f thiyo s proces s gooi competitivsd an d n averagi e e with fossil-fuelled sources. As base sources for centralized district heating and warm water supplying systems the nuclear CHP sources are well suite r lonfo dg heat delivery periods. With the specialized nuclear heating plants (NUP) not too much experience has yet been gained. If we omit the Swedish decommissioned Agesta NHP, the e onle nowadayth n on y s operatin ge Canadia th NH s i P n 2MW SLOWPOKE Energ piloo y tw System te plantTh . s AST-50 e USSth e ar Rn i 0 still under construction. Generally this typ f nucleao e r heat sources i s therefore not yet approved in operation and also the economy cannot yet be well defined. From the technical point of view this type of nuclear installations is very perspective because already the designs of reactors for heating plants brought several important new design approaches and in the first range the "Inherent-safety" principles. The employment of "natural physical laws" in self regulation and control processes is now widely envisaged in the advanced reactor concepts. It can be therefore expected that - even if later than was foreseen - the MIPs will be applied s safd reliabla an e e source r producinfo s g hear spacfo t e heatin ward an gm water supply. The expected time horizon could be after the year 2000. The IAEA will in the future look into questions of nuclear heat applicatio n broadeni r context wit e smal mediuth d h an l m power reactors and advanced reactor systems. 65 Annex 1 MAIN DATA OF NUCLER HEATING PLANTS COUNTRY : USSR FRC SWEDEN FRANCE SWITZERLAND CANADA CHINA FINLAND People1! Republic AST- 500 AST-300 U W HEREK - SECURE THERMOS SHR GHR SLOWPOKE NHR 500 200 300 4OO 200 100 150 t MW 5 ES Typ f reactoo e r R PW R PW BWR R PW R BW BWR R PW PWR R PW R BWR CCR PWR PWR Thermal power MW 500 500 0 30 30 00 20 0 50 •5 400 0 0 201 15 0o t 0 2 10 0 1 0 1 Coo/ /modért n l a 0 2 r H o at 0 2 H / 0 2 H H20 H20 / H20 H20 /0 2 HH 2 0/ 0 2 H H20/H20 He/C H20/H20 r^O/h^O Primary circuit Type of design integr. integr. integr. 2 loops integr. integr. integr. tank integr. Circulation NC NC NC C N FC C N FC C F NC Core temperature outleC t 200/150 200/120 198/158 120/90 144/131 198/185 93/68 198/? inlet C° 208/131 198/163 165/115 115/90 144/125 450/250 Pressura MP e 0 2, 0 2, 1,6 1,5 1,5 1,5 0,7 0,7 1,3 1,5 5 1, 1,0,15 5 71, % sceam quality 0 0 2 o t p u mass 0,9 0,9 0 0 0 0 0 0 - 0,8 Intermediate circuit Temperature ° C delivery 170/90 130/75 165/100 137/96 180/125 85/60 return 160/90 160/125 153/75 127/91 135/95 Pressura MP e 1,2 1,2 1,2 1,7 1,7 1,7 1,6 1,6 1,7 0,8 0,1 Heating system circuit TemperaturC ° e delivery 150/70 144/64 120/60 120/70 120/70 140/60 100/70 100/70 130/80 120/80 120/60 120/60 80/55 Pressure MP0 2, a 0 2, 2,6 1,0 8 0, 0,8 1,30,7 1,4 1,6 1,5 0,1 1,5 0,8 1,7 oo COUNTRY : G FR SWEDEN USSR FRANCE SWITZERLAND CANADA CHINA FINLAND People's Republic AST-500 U W AST-30 K HERE- 0 SECUR R E GH R SLOWPOKNH R THERMOSH E S t MW 5 S E 0 15 0 10 0 20 0 40 0 30 0 20 0 50 Type of reactor R PW BWR PWR BWR BWR PWR PWR PWR PWR PWR BWR GCR PWR PWR Core Height/ 3000 3000 2920 2350 3807 1974 1200 - 800 , 3 495 580 diameter, mm 2800 2800 3300 3136 1760 1200 964 5m 306 600 Assembly design hexag.hexag hexag. square square square square square square - square pebble square square bed Numbef o r 2 1 4 0 35 8 2 2 3 2 3 4 assemb14 s 8 lie 28 4 14 1 12 120 1 18 8 36 85 Numbe f lato re tic "Caramel" 9 4 position/ass. 0 8 6 typ7 e 0 fue6 168l 16 8 824 4 6 4 6 Fuel element design l cy c. y1 cyl. cyl. cyl. cyl. cyl. cyr. plate cyl. spheric cyl. cyl. Core power 5 5 27,5 2 7 4 5 1 densit4 y KW/1 2 1 27, 0 2 4 30 0 2 23,0 Nuclear fuel Mass, tU 50 50 54 20 70,6 26 13 2,7 1,2 0,34 0,111 0,453 Enrichmenf o t refuel ling material 7. 1,6;2,0 l,6;2,0 5 1,23;2,52 2,7 2,58 3,7 4,5 20 5 3 Refuelling fraction 1 1/3 1/4 3 1/ 1 1/3 .4 1 ..1/1/ 4 1 1 Fuel reloading cycle à 2 2 20 20 2 2 2 12 16 3 COUNTR : Y G FR SWEDEN USSR FRANCE SWITZERLAND CANADA CHINA FINLAND People'« Republic AST-500 U W AST-30 K HERE-0 SECUR ER GH SLOWPOK R THERMOSH E S NHR S E 0 15 0 10 0 20 0 40 0 30 0 20 0 50 5 MWt R PW R GC R BW R PW R PW R PW R PW R PW R BW R BW R PW R PW Typ f R reactoo eBW r PWR Residence time à 6 6 8,8 20 20 88 6. ..8 12 16 3 Average burnup MWd|tU 15000 14000 40000 14000 27000 22000 30000 17000 85800 11200 4600 Fuel power density , MW|tU 10 10 8,3 10 4,25 15 15 37 8,1 18 11,3 - natura C N l circulation FC - forced circulation O\ Annex2 USSR SITING REQUIREMENTS FOR NUCLEAR DISTRICT HEATING PLANTS ANNUCLEAR DFO R CO-GENERATION WITH REGAR RADIOLOGICAO DT L SAFETY* (AnnexGeneralthe to Regulations Safety the Nuclear of for Power Plants During Design. Construction and Operation) 1. GENERAL REGULATIONS 1.1. Nuclear district-heating plants (nuclear reactors) may be built at least 2-3 km from projected boundaries of densely populated urban centres' provided thaadditionae th t l safety requirements give Section ni met.e ar n2 1 1.2. When siting a nuclear district-heating plant the following radiation limits for exposure of the population should not be exceeded: During normal plant operation the maximum individual radiation dose populatey inan d area shoul t exceedno mrem/year0 d2 , excepr fo t exposur thyroie th othef d eo dan r critical organs, wher dose eth e limit mrem/yea0 is6 thyroie th r childrenf fo rd o , whil collective eth e dose should not exceed 104 man-rem/year for the total population; For design basis accident conditions the maximum individual radiation dose outside the perimeter fence of the plant should not exceed 10 rem, excep exposurr fo tthyroi e th f eo f children do , whil collective eth e dose shoul t exceedno s man-re10 d m unde worse rth t weather conditions. 1.3. Nuclear co-generation plants with reactor facilities which confore th o mt existing General Regulationsan3 d whose specifications mee requiree th t - ments given in paragraph 1.2 on radiation limits for town populations for nuclear district-heating plants may be used as sources for district heating. • Ratified by the USSR Ministry of Power and Electrification, by the USSR State Committee on the Use of Atomic Energy, and by the USSR Ministry of Health in their Resolution of 5 October 1978. projectee Th 1 d boundar towne th regarde s f i yo dthau t containe approvee th dn i d develop- ment plan of the town. Further development of the town will need to take into account the existence of the nuclear district-heating plant and the number of years of its operation. 1 Applicatio additionae th f no l requirement Nortr Section s i Fa r region r he fo fo nth 2 n si sparsely populated centre treates si d separatel eacr yfo h specific case. 1 General Regulation Safete th Nuclear f yo sfo r Power Plants During Design, Constructiod nan Operation, hereafte thin i r s paper referre Generas a o dt l Regulations. 71 These nuclear heating site e plantb y d withisma nzona f denseleo y popu- lated ce/itret closeno t r sbu tha distancee nth s fro projectee mth d urban boundaries whic tabulatee har d belourbae sizth relate d f wth nean o o et population4: Population density 100-300 300-500 500-1000 1000- 2000 (in thousands) oved an 200r 0 Distance to the 10 12 18 25 boundar) km n y(i l typeAl 1.4f nucleaso . r district-heating plant must mee requiremente th t s in Sectio whicn3 h prohibit radioactive substances reachin consumee gth r throug heatine hth g supply network. ADDITIONAL SAFETY REQUIREMENT NUCLEAR SFO R DISTRICT-HEATING PLANTS Measures should be envisaged to prevent core melt-down in the event of damage to any of the reactor vessels leading to leakage within the permissible limits determined by the design features of the vessel. The impossibility of large-scale leakages should be demonstrated. 2.2. The installations and systems of the plant should be laid out and designed to allow for external events such as an aircraft crash, explosions in neigh- bouring industrial plants, or in passing vehicles, etc. In such cases the population exposure limits, definemaximue th r dfo m design basis accident, should not be exceeded. 2.2.1 e desigTh . n parameter aircrafn a r sfo t cras mase h ar tonnes 0 s2 , velocity 700 km/h, with the load applied over an effective cross-section of 7 m2. After the crash the possibility of its fuel catching fire must be considered. 2.2.2 desige Th . n parameter shocr sfo k wave maximua e sar kg/cm5 0. f mo 1 over a period of 10 seconds. 2.2.3. In the event of an explosion or aircraft crash at least one channel of the defence systembarriee on d r channero s an acciden e th n i l t confinement systems should still be able to operate. desige th t nA 2.3 stage .permissibl e th , e leve f leakag o l leaktighe th n ei t contain- ment system should be established and it should be proved that this level cannot be exceeded. The degree of leaktightness obtained during design should be confirmed after manufacture and regularly checked during operation. 2.4. Provision should be made for the removal of liquid and solid radioactive waste of different levels of radioactivity and for interim storage (up to five years site n als d )o reprocessinr o,an fo solidificatiod gan f wastno siten eo . r townFo s wit populatioha morf no e thimillioo ntw n peopl sitine nucleaa eth f go r co-generatio4 n plant 'a considered separately for each specific case. 72 2.5. The technological systems for treating radioactive waste should ensure that it is sufficiently decontaminated to be re-used in the technical water supply. The total annual activity of the liquid waste released must be established for each desige planth t a nt stage. operatine Th 2.6. g condition fabricatioe sth for d ,an , fuenof l elements should guarantee tha t exceet theno damage o yd th e limits established: 0.1% loss of leaktightness and 0.01% contact between the coolant and fuel elemen r scinterefo t d uranium dioxide equivalene th r o , t releas f activiteo y into the circuit for other types of fuel. 2.7. Before commissioning a nuclear district-heating plant the licensing authorities should approv repore eth "Technican o t l safet operatiod yan n of the plant", corrected on the basis of results obtained during adjustment and com mission ing tests. This repor prepares i t operatine th y db g organization together wit desige hth n organization scientifie th , c directoheae th dd designeran r . 3. REQUIREMENTS FOR PREVENTING RADIOACTIVE PRODUCTS FROM ENTERING THE HEATING SYSTEM heatine Th 3.1g. network shoul designee db preveno dt possibilite th t f yo radioactive products entering the system. The following minimum requirement essentiale sar . 3.1.1. Heat from the reactor coolant should be removed via leaktight heat-transfer surfaces to an intermediate thermal medium. 3.1.2. Heatin heatine th f go g networ thermae th y kb l medium must take place only via an intermediate surface. 3.1.3. The pressure of the thermal medium should be lower than the pressure of the heating network. 3.2. Both during normal operating conditions and in accident conditions the radioactivity of the thermal medium should not be more than 10 times the permissible concentration limit r watesfo r give documenn ni t NRB-76 (Basic Safety Standards for Radiation Protection). 3.3. The heating network should be isolated from the heat exchanger with the thermal medium in the event of accidental penetration of radioactive substances int heatine oth g network which might buildue resulth n i t p of radionuclide excesfollowinn e si th f s o limitso gtw time1 0. :e sth permissible concentration level for water given in NRB-76 and 10 times the permissible concentration level of radioactivity in the water supply. heat-exchangere Th 3.4. district-heatine th r sfo g system shoul locatee db n do the site of the plant. 73 LIST OF PARTICIPANTS IAEA Podes. M . tMr International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria CANADA Mr. FentoN . n Atomic Energ f Canado y a Limited Whiteshell Nuclear Research Establishnent Tel.: 20A-753-2311 Telex: 07-57553 CHINA, Peoples Republic of u H O ZU . Mr Bureau of Science and Technology Ministry of Nuclear Industry g Bein ij Telex: 22240 CNEIN C CZECHOSLOVAKIA Mr. Josef Kadlec Research Institut f Fueo ed Energ an l y Complex, VUPEK, Vladislavova 4 11372 Prague 1 Tel.: 741351 ext. 453 Telex: S 121Vd 76 2 Mr. J. Lhota Energopro ject Bubenska 1 17005 Pragu7 e Tel.: 879583 Zlatnansk. J . Mr y Federal Ministr f Fueo yd Powe lan r Vinohradska 8 12070 Prague 2 Tel.: 8 (0042217 3 1 3 )2 3 Telex08 2 :12 FINLAND Mr. T. Haapalainen Imatran Voima Oy Power Station Department Eerikinkatu 27, Box 138 00101 Helsink0 1 i Tel.: (003580) 61601 Telex: 124608 voimf s a 75 FRANCE Mr. CaizergueR . s CEA Sac lay DEMT/SEKMA Bâtiment 30 91191 Gif-sur-Yvette CEDEX Tel.: 69.08.55-14 GERMANY, Federal Republif o c Mr. Dietmar Bittermann KVriJ/R 4 Erlange92 , n Hammberbacherstr. 12-14 Tel./FAX.- (09131) 18-0 Telex: 62929-0 Mr. R. Rotterdam Hochtemperatur Reaktorbau GmbH Postfach 5360 6800 Mannheim1 Tel.: 0621 451 483 GERMANY, Democratic Republic of Mr. L. Ackermann Staatliches Amt fuer Atomsicherheit und Strahlenschutz DDR 1157 Berlin Waldovalle7 11 e Tel: 5020 Telex: 112632 SAAD D S Mr. LuetzoK . w Ingenieurhochschule Zittau DDR 8800 Zittau Theodor-koerner-Allee 16 Tel.0 61 : HUNGARY Mr. MorentA . h Ministr f Industro y y H-1525 Budapest Pf. 96 Tel.: 558-372 SWITZERLAND . MrFoskoloG . s Swiss Federal Institut r Reactofo e r Research CM-5303 Wuerenlingen Tel.:1 1 (056 1 2 9 )9 Telex:H C 5371 r ei 4 UNION OF SOVIET SOCIALIST REPUBLICS Mr. V.A. Ivanov GOSATOM Moscow Mr. I.N. Sokolov GOSATOM Moscow 76 V.S. Hr . Kuul GOSATOM Moscow Mr. A.F. Osokin GUSATOM Moscow OBSERVERS (al f theo l m from Czechoslovakia) Mr . Baraba.K s Czech Technical University Suchbatarova Prague 6 . MrKlaiM . l Research Institut f Fueeo d Energ an l y Complex, VUPEX Vladislavova A 11372 Pragu1 e Mr. kliF . k Czech Technical University Suchbatarova Prague 6 Mr. M. Novotny East Bohemia Electric Concern Opatovice n. 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