IAEA-TECDOC-436

GAS-COOLED REACTORSAND THEIR APPLICATIONS

PROCEEDINGS OF A TECHNICAL COMMITTEE MEETINGON GAS-COOLED REACTORS AND THEIR APPLICATIONS

ORGANIZED BY THEINTERNATIONAL ATOMIC ENERGY AGENCYAND HELD IN JÜLICH, 20-23 OCTOBER 1986

A TECHNICAL DOCUMENT ISSUED BY THEINTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1987

GAS-COOLED REACTORS AND THEIR APPLICATIONSIAEA, VIENNA, 1987IAEA-TECDOC-436

Printed by the IAEA in AustriaOctober 1987

FOREWORD

Technological developments for gas-cooled reactors are at presentbeing carried out on a large scale in the UK for C(>2-cooled AdvancedGas-Cooled Reactors (AGRs) and for High Temperature Gas-Cooled Reactors(HTGRs) in the Federal Republic of Germany, United States of America,Japan and USSR.

All programmes are directed towards construction of plants. Theprogramme in the UK is by far the most advanced effort with severalcommercial Hagnox Reactor and Advanced Gas-Cooled Reactor units inoperation and further AGRs in the construction or commissioning phase.HTGRs are in operation in the USA and in the Federal Republic ofGermany. Other HTR specific work is being carried out in Austria, Chinaand Switzerland. At the IAEA the International Working Group onGas-Cooled Reactors (IWGGCR) has been established in 1978in order to promote an exchange of information on gas-cooled reactordevelopment and to stimulate international cooperation. In the frameworkof this working group, other countries are following the development ofGCR technology in order to investigate its possible benefits for nationalenergy supply. The IWGGCR has recommended to the Agency to convene thisTechnical Committee on Gas-Cooled Reactors and their Applications, whichwas attended by more than 200 participants from 25 countries andInternational Organizations.

The Agency is grateful to the Government of the Federal Republic ofGermany and to the Nuclear Research Centre Jiilich for their hospitableand efficient arrangements.

This volume contains all papers presented at the meeting.

EDITORIAL NOTE

In preparing this material for the press, staff of the International Atomic Energy Agencyhave mounted and paginated the original manuscripts as submitted by the authors and givensome attention to the presentation.

The views expressed in the papers, the statements made and the general style adopted arethe responsibility of the named authors. The views do not necessarily reflect those of the govern-ments of the Member States or organizations under whose auspices the manuscripts were produced.

The use in this book of particular designations of countries or territories does not imply anyjudgement by the publisher, the IAEA, as to the legal status of such countries or territories, oftheir 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 anyendorsement or recommendation on the part of the IAEA.

Authors are themselves responsible for obtaining the necessary permission to reproducecopyright material from other sources.

Please be aware that all the Missing Pages

in this document were originally blank pages

CONTENTS

Introduction and Summary ................................................................................... 7

OVERVIEW OF THE STATUS OF GAS-COOLED REACTORS ANDTHEIR PROSPECTS (Session A)

Magnox reactors and advanced gas-cooled reactors .................................................... 11A.F. Pexton

High temperature gas-cooled reactors ..................................................................... 29R. Schulten

EXPERIENCE WITH GAS-COOLED REACTORS (Session B)

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

Technical evolution and operation of French CO2 cooled reactors (UNGG) ...................... 53Y. Berthion

Fort St. Vrain performance ................................................................................. 61H.L. Brey

AYR experience ............................................................................................... 71E. Ziermann

THTR operation — The first year ......................................................................... 99D. Schwarz

DESCRIPTION OF CURRENT GCR PLANT DESIGNS (Session C)

Restoration of a CEGB Magnox reactor to full power ................................................ IllR.J. Paynter, B.J. Roberts, P. T. Sawbridge

The modular high-temperature gas-cooled reactor (MHTGR) ........................................ 123A.J. Neylan

Status of gas-cooled reactor development in the USSR ................................................ 135V.N. Grebennik

The nuclear electricity and heat generation using the VG-400 reactor ............................. 141V.V. Bulygin, V.E. Vorontsov, V.N. Grebennik, E.M. lonov, A.N. Protsenko,A.Ya. Stolyarevskij, S.V. Shishkin, Yu.K. Panov

Current status of research and development of HTGR in Japan ..................................... ' 51T. Hayashi

Plant design of a high temperature engineering test reactor in JAERI ............................. 165O. Baba

Status of the HTR-Module plant design .................................................................. 179I.A. Weisbrodt, W. Steinwarz, W. Klein

Small nuclear power plants : 10 M W GHR gas-cooled heating reactor ............................ 199H. Schmitt

HTR 100 industrial nuclear power plant for generation of heat and electricity ................... 213S. Brandes, W. Kohl

HTR 500 — The basic design for commercial HTR power stations ................................ 235E. Baust, J. Schöning

SAFETY ASPECTS (Session D)

Safety characteristics of current HTR ..................................................................... 253W. Kroger

The safety concept of the modular HTR ................................................................. 267G.H. Lohnen

MHTGR licensing approach and plant response to off normal events .............................. 283A.J. Neylan, F.A. Silady

Comparison of regulatory aspects in different countries ............................................... 295K. Hofinann

GAS-COOLED REACTOR APPLICATIONS (Session E)

HTGR applications, prospectives and future development ............................................ 313H. Barnert

Perspectives of HTGRs in chemical and iron and steel industries of Japan ....................... 329K. Tsuruoka, T. Katamine, T. Miyasugi, R, Araki

Israeli perspective on HTGRs ............................................................................... 343A. Barak, A. Beck, E. Greenspan, J. Szabo, L. Blumenau, H. Branover,A. El-Boher, E. Spero, S. Sukoriansky

GAS-COOLED REACTOR TECHNOLOGY (Session F)

Overview of US MHTGR base technology development program .................................. 371J.E. Jones, Jr., P.R. Kasten

Status and development of the German materials programme for the HTGR ..................... 387H. Nickel, F. Schubert, H. Schuster

KVK and status of the high temperature component development .................................. 407W. Jansing, H. Breitling, R. Candeli, H. Teubner

HTR fuel development and qualification — Treatment of spent fuel ............................... 429G. Kaiser, K. Hackstein

Characteristics of HTGR spherical fuel elements ....................................................... 445A.S. Chernikov, L.N. Permyakov, L.I. Mikhajlichenko, S.D. Kurbakov

Self-sustaining thorium cycle in a high temperature graphite reactor ............................... 461K. Balakrishnan

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

Future applications of the high temperature reactor .................................................... 473A. Klusmann, M. Stelzer

Financing models for HTR plants — Co-finaning, counter trade, joint ventures ................. 483J. Bogen, D. Stölzl

Perspectives on the HTGR from utilities in the USA .................................................. 499H.L. Brey, L.D. Mears

Market prospects of HTRs hi newly industrialized and developing countries ..................... 511S. Garriba, C. Vivante

Chairmen and Secretariat .................................................................................... 521List of Participants ............................................................................................ 523

INTRODUCTION AND SUMMARY

On the invitation of the Government of the Federal Republic ofGermany the International Atomic Energy Agency convened in Jiilich,20-23 October 1986, the Technical Committee on Gas-Cooled Reactors andtheir Applications. The meeting was hosted by the Nuclear ResearchCentre Jiilich and cosponsered by the Commission of the EuropeanCommunities. The purpose of the meeting was to review and discuss thecurrent status and recent progress made in the technology and design ofgas-cooled reactors and their application for electricity generation,process steam and process heat production. The meeting was attended bymore than 200 participants from 25 countries and InternationalOrganizations presenting 34 papers.

The technical part of the meeting was subdivided into 7 sessions:A. Overview of the Status of Gas-Cooled Reactors and their ProspectsB. Experience with Gas-Cooled ReactorsC. Description of Current GCR Plant DesignsD. Safety AspectsE. Gas-Cooled Reactor ApplicationsF. Gas-Cooled Reactor TechnologyG. User's Perspectives on Gas-Cooled Reactors

At the end of the meeting a round table discussion was organized inorder to summarize the meeting and to make recommendations for futureactivities.

Gas-Cooled Reactors have been under development for more than threedecades. Today there are more than 40 gas-cooled reactors in operationor in the commissioning phase in seven Member States of the IAEA. Mostof them are C02~Cooled Magnox Reactors and Advanced Gas-Cooled Reactors(AGRs) for electricity generation, having accumulated about 800 reactoryears of operation. The Helium-Cooled High Temperature Gas-CooledReactors (HTGRs) are also for electricity generation, however, the longterm development goal for the HTGR is in most countries the productionand application of process steam and process heat for chemicalindustries, coal conversion processes, nuclear steelmaking, etc.Currently there are also gas-cooled reactors of small sizes under designwith reduced complexity, passive safety mechanisms and a high potentialfor modular design and shop fabrication. These reactors are not onlyappropriate for application in highly industrialized countries but alsoin countries with less developed industrial infrastructure.

During the meeting the technology of gas-cooled reactors waspresented and discussed together with their different applications.Particular emphasis was given to a review of their safety features, and

the perspectives of users on this reactor line. After the meeting a tourtook place to the THTR-300, a 300 MWe pebble bed reactor, which is in thecommissioning phase and which has successfully reached the 100% powerlevel.

The meeting concluded that gas-cooled reactors have reached a veryhigh level of maturity. In the UK the AGRs, in particular theHinkley/Hunterston type design, have proven to be economic and safeelectricity producers. The successful operation of High TemperatureGas-Cooled Reactors in the USA and the Federal Republic of Germany andthe commissioning of the THTR-300 have proven that this technology isavailable for electricity generation and process steam production. Theadvantages of the system, i.e., high safety, high temperature potentialand diverse applications, environmental acceptability and fuel savinghave been confirmed by the operation of prototype and demonstrationplants. GCRs are a unique tool for power and steam production, a toolwhich is based on the margins of safety inherent in the design conceptrather than on engineered safeguards, a feature which is of increasinginterest in Agency Member States. Because of their specific properties,GCRs are expected to have very good international market chances inparticular in the medium sized and small power range.

OVERVIEW OF THE STATUSOF GAS-COOLED REACTORS AND THEIR PROSPECTS

(Session A)

MAGNOX REACTORS AND ADVANCEDGAS-COOLED REACTORSA.F. PEXTONSouth of Scotland Electricity Board,Glasgow, United Kingdom

Abstract

Commercial nuclear power in the U.K. is based on CC>2 cooledgraphite moderated reactors. This started around 1950 with themagnox programme having Calder Hall and Chapelcross as theforerunners. Nine twin-reactor commercial magnox stations werethen built between 1958 and 1970. The first seven stationshave steel pressure vessels with external boilers connected bygas ducts. The last two have an integrated design with core,boilers and gas circulators enclosed in pre-stressed concretepressure vessels. The. magnox reactors have given good serviceand the older stations are now being subjected to careful reviewafter 20 years operation.

Magnox was superseded by the AGR and five twin reactor stationsto three separate and distinct designs, but all with integralgas circuits inside concrete pressure vessels, were ordered inthe period 1965 to 1970. The most successful of these, atHinkley Point 'B' and Hunterston 'B' have now been in operationfor over 10 years. Their load factor for the past few yearshas been of the order of 80%.

In 1979 two further twin reactor AGR stations, Heysham II andTorness, were authorised. These were based on the Hinkley/Hunterston design with detailed improvements arising fromoperating experience and developments in safety standards.Commissioning tests are now well advanced and the first reactorsat each station are scheduled to come into operation in 1987.Work has proceeded close to the original programme and budget.The Heysham II/Torness design now provides a standard base foron-going development of the AGR. A study is in hand whichessentially replicates the main features, but by effective useof margins enables higher output to be achieved.

The future depends on government policy following issue of thereport of the inquiry into CEGB's application to build a PWR atSizewell, as well as on the political environment followingChernobyl. Whatever the outcome, however, the AGR provides anattractive option with many inherently favourable safetycharacteristics. Whilst it is difficult to see an incentivefor the U.K. to change to an alternative form of gas-cooledreactor technology, U.K. gas cooled reactor experience providesscope for transfer of technology to HTGR developments in Europeand the U.S.

11

The Structure of Electricity Supply

In the UK, production and distribution of public electricitysupplies is the responsibility of State-owned monopolies. Thereare, for planning and commercial purposes, two integrated gridsystems, one for England and Wales and one for Scotland but theinterconnection capacity between them, some 1000MW, is adequateto allow for power exchanges arising from economic scheduling ofthe generating plant on both sides of the border.The Contribution from Nuclear Power

The nuclear stations serving the overall U.K. grid system areshown in Figure 1.

Dounrcay

Hmkl«» Point A S Winfnth

StationBerkeleyBradwellCalder HallChapetcrossDungeness AHinkley Point AHunterston ATrawsfynyddSliewellOk) buryWytlaHinkley Point BDungeness BHarttepoo)Heysham 1Heysham II*Hunterston BTomess*OounreayWlnfrrth

•not )r*t commlMàorwd

TypeMagnoxMagnoxMagnoxMagnoxMagnoxMagnoxMagnoxMagnoxMagnoxMagnoxMagnoxAGRAGRAGRAGRAGRAGRAGRPFRSGHWR

Output(MWso)

276245200200424430300390420434840

1040330033003300123011501360250100

FIG. 1. UK NUCLEAR POWER STATIONS.Commercial nuclear power is based on CC>2-cooled, graphitemoderated reactors, starting with the Calder Hall 'Magnox' type,which uses metallic natural uranium fuel with magnesium alloycladding. Eight twin reactor commercial plants of this type,all to different, progressively larger and more developeddesigns, were built by The Electricity Supply Industry inEngland and Wales, and one station, Hunterston 'A' was built inScotland, all between 1958 and 1970, providing a total of4000MW(e). At one time there were five different industrialconsortia competing for the design and construction of nuclearpower stations.

The commissioning of a 33MW(e) test bed for the Advanced Gas-Cooled Reactor (AGR) at Windscale in 1962 opened the way twoyears later to an evaluation of competitive tenders for AGR, PWRand BWR. This led to adoption of the AGR system. Fivestations, each with twin reactors of some 600MW(e), were orderedin the period 1965 to 1970, to three separate and distinctdesigns from the three nuclear design and construction companies

12

still active in the UK. Two more twin reactor stations, CEGB'sHeysham II and SSEB's Torness were authorised in 1979 based onCEGB's Hinkley Point 'B' and SSEB's Hunterston 'B', the mostsuccessful of the three original designs.II and Torness is scheduled in 1987.

Start-up of Heysham

At present nuclear power provides 19% of the total energygenerated for public electricity in the UK but this will grow,with further plant now building and being commissioned, to morethan 25% in the next few years. In Scotland the nuclear shareis much higher, the present 45% increasing to over 60% withinthe next two years. These outputs are compared with that ofother countries which have a significant nuclear contribution inFigure 2. Actual information is shown for 1984 together withthe OECD estimates for 1990.

FIG. 2. NUCLEAR SHARES OF ELECTRICITY PRODUCTION.

The annual requirement for increased capacity in the UK isestimated at about 1200MW(e) per year (ie one plant per year)from the mid 1990s. In Scotland alone the correspondingrequirement would be only one 1200MW(e) plant per decade but, inview of the long timescale required for planning and approvals,thought is already being given to the next plant.

Based on current cost estimates the advantages of AGR comparedwith new coal-fired plant for base load operation is at least30%, expressed in total cost per unit sent out, and could bemore depending on the extent of the provisions required for fluegas desulphurization and low production of nitrogen oxides.Clearly in the UK as a whole there is a strong economicincentive to increase the proportion of nuclear in the plantmix. However, if the full benefits of a nuclear ordering

13

programme are to be achieved, it will be important that thisrests on the twin foundations of a reactor design which can bereplicated without significant changes and a constructionprogramme in which there is sufficient assurance of successiveorders to allow proper planning of resources and careerstability for trained and experienced personnel in the design,construction and manufacturing sectors.The Magnox Programme

Calder Hall, in England, and Chapelcross, in Scotland, the fore-runners of the magnox programme each comprise four reactors of35 MW(e) and operate at 7 bars gas pressure. Each reactor hasfour separate gas circuits, each containing its own gascirculator and boiler (Fig.3). Refuelling is carried out offload. The fuel element cans originally had simplecircumferential fins but these were later modified to aherringbone design which gave improved heat transfer. This,with other improvements, enabled the net output to be increasedto 50MW(e) per reactor and the uprating and the historical loadfactor of Chapelcross are shown in figure 4. After 25 years ofoperation, the lifetime load factor of the station, based onrated output, is 85%.Hunterston 'A' was the third twin reactor commercial magnoxstation built in the U.K., and its first unit was commissionedin 1964. Each reactor has eight separate gas circuits withconfiguration shown in figure 5. The carbon dioxide pressure is10 bars and the design net output is 150Mtf(e) per unit.Refuelling is carried out at full load using a refuellingmachine having access to pressure vessel standpipes below thereactor. The historical load factor of Hunterston 'A' is shownin figure 6. After 22 years of operation it has a lifetime loadfactor of 82% based on design output.

iZSâV/////^//////. mm^

FIG. 3 CROSS SECTION THROUGH CHAPELCROSS REACTOR BUILDING14

200 _

MTtDOUTPUT

100

...r

X100—-

rtcrai

N.

u.

ij970'7t ' n ' TJ ' 74 ' n ' n ' n ' nT n 'IMO' it ' R ' n ' n ' « 'YEARS

>gBO > (t ' S ' O ' 64 ' «5 ' * ' «7^ H l H 'iPO1 7l1« ' 7 J ' 7 4 l 7 S l n l 7 7 l 7 i ' 7 » l » l H ' i » l « » l i 4 l «YEARS

FIG . A ANNUAL LOAD FACTORS OF CHAPELCROSS BASED ON RATED OUTPUT

CHARGEMACHINE

GASCIRCULATOR

FIG 5 HUNTERSTON "^ MAGNOX

15

MWSO

DESIGN

OUTPUT

300

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

100 -T-

90 --

LOAD

FACTOR

ten.

FINANCIAL YEABS /aiFIG . 6 ANNUAL LOAD FACTORS OF HUNTERSTON 'A' BASED ON DESIGN OUTPUT

The first seven U.K." commercial magnox stations have steelreactor pressure vessels with external boilers connected byducts. A major change was introduced in CEGB's last two magnoxstations, Oldbury and Wylfa, which have an integrated designwith the whole of the gas circuit, including core, boilers andgas circulators, enclosed in a prestressed concrete pressurevessel.In 1969 a gas circuit corrosion problem was encountered whichaffected all magnox stations. The major structural steelcomponents had been manufactured from silicon killed mild steelwith good corrosion resistance in moist CC>2 but certain otheritems had been made of more rapidly corroding rimming steel.The growth of oxide on the faces of corroding washers, forexample, had in some areas built up local strains sufficient tofail bolts. Derating of some of the reactors was necessarywhilst a systematic programme of monitoring, development andmodification was undertaken. At Hunterston 'A', for example,the mean duct outlet temperature was reduced from 390°C to 360°Cgiving a fall in net station output from the design value of300MW(e) to less than 280MW(e). Flow tests on a scale model ofthe reactor demonstrated that peak temperatures in the mostcritical areas could be reduced by leaving a few peripheral fuelchannels empty. As a result the output has now been restored tothe original design value.Magnox fuel is a success story. The design burn-up ofcommercial magnox fuel was initially 3000MWd/Te, limited by theexpected physical life of the fuel elements. The average

16

channel burnup now being achieved is over 5,500MWD/Te. Over2.5 million fuel elements have now been loaded into thecommercial reactors with an average failure rate less than 0.1%,and the current failure rate is an order of magnitude less thanthis.The Nuclear Installations Inspectorate decided that after 20years operation that the magnox plants should be subject to acomprehensive safety review. The entire range of safety hasbeen studied including structural integrity of the pressurecircuit, core and components, the ability of the reactorprotection to respond to all types of reactor faults and theability of structures to withstand external hazards. OnHunterston 'A' alone this review has absorbed over 100 man yearsof professional effort.The AGR ProgrammeThe commercial AGR programme adopted the concept of the integraldesign inside a concrete pressure vessel.All five of the AGR stations ordered between 1965 and 1970 arenow operational and showing encouraging potential. HinkleyPoint 'B' and Hunterston 'B' have been generating for over tenyears. Figure 7 compares cross sections of Hunterston 'B' andthe subsequent design development at Torness.

SHIELDINGSA5 BAFFLE

CONCRETE PRESSURE VESSELREACTORBOILERDECAY HEAT BOILERGAS CIRCULATOR

SECONDARY SHUTDOWN ROOM

HUNTERSTON 'B1 TORNESS

FIG.7 COMPARISON OF REACTOR PRESSURE VESSEL CROSS SECTIONSFOR HUNTERSTON 'B' AND TOHNESS

17

The rated output of each unit at Hunterston was originallylimited to 460 MW primarily by uncertainty as to the corrosionresistance of the 9% Cr. steel boiler tubes and by some minordeficiencies in fuel pin end-caps. With the progressiveintroduction of detailed improvements the net authorised ratinghas been raised to 575MW(e), close to the original designtarget. Further increases will be made as modified blades arefitted to certain of the turbine stages and as improved Stage 2reactor fuel elements are introduced.

The load factor has been progressively improved, first byeliminating sources of unreliability (mainly in the conventionalplant) and, secondly, by establishing on-load refuelling asnormal practice. It is necessary to reduce power to 30 per centfor refuelling at present to avoid the risk of damaging thegraphite sleeves of the fuel assemblies but the Stage 2 fuel hasa more robust graphite sleeve and will allow the refuellingpower to be raised. Further reductions of refuelling outputlosses will therefore be achieved as Stage 2 fuel comes to bedischarged.Figure 8 shows progress in raising the output capability andload factor at Hunterston. Achievement at Hinkley has beensimilar.Construction of the earlier AGRs entailed fabrication on site ofsuch large components as the core supports, the gas baffles andthe core restraint tanks. However, for Heysham II and Torness,having established an agreed national design, it was decided tosecure the benefits of factory manufacture and these componentswere transported fully assembled to site. New manufacturingfacilities were created to produce not only these largefabrications but many others such as graphite blocks, gascirculators, boilers and fuelling equipment, all to very high

10CU90_BO-

LOAD 70-FACTOR--

% 60-50_40_30_ao_10-0

1977/78

1

1978/79' 1979/80

i —————

19BO/B1 1981/88

1

198a/83 1983/84

, _____

'

1984/83 19B5/8É

r-12004100-1000_900BOO

-700 DECLARED.600 NETT

CAPABILITY-500 MW.400_300-200_100

FINANCIAL YEAR

FIG.8 CAPABILITY AND LOAD FACTOR FOR HUNTERSTON B.

18

standards of quality and with good productivity. Necessarilysite-performed operations, such as thermally insulating thepressure vessel, were rationalised using the most up-to-datetechniques and full-scale mock-up training facilities wereprovided for the operatives at site. Figure 9 shows the siteconstruction sequence employed with the large prefabricatedcomponents.

PRESSUREVESSELLINER \

ATTACHEDPENETRATIONS

GAS BAFFLE

PHASE 1ROLL-IN PRESSURE VESSEL LINER

PHASE 2LIFT-IN GAS BAFFLE ASSEMBLY FOLLOWED BYINSTALLATION OF PRESSURE VESSEL ROOF

LECTRICAL ANDINSTRUMENT•AREAS

PHASE 3MAIN MECHANICAL/ELECTRICAL WORK FACES

F16.9 A6R CONSTRUCTION PHASES AT TORNESS

Figure 10 shows the achievement in construction of the two unitsat Torness against the original programme set in 1979. Theconsent of the Nuclear Installations Inspectorate is expectedshortly for the loading of fuel in Reactor 1. The unfuelledcommissioning tests have established the integrity of thepressure vessel and the gas circuit and other components of theplant have behaved as designed during hot pressurised gas flowtests.Hunterston 'B' was a prototype and the many lessons learned fromits construction and operation have been incorporated in theTorness project. Measurements of actual operating performanceat Hunterston have allowed the rated output of Torness to beincreased by 7 per cent without plant modifications. Once the

19

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UNIT 2

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BAFFLE

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JUL 1986

COMPLETEBOILERHYDRAULICTESTSUK 1986

SYNCWONISEUNIT 2

MAR 1988CONBtHB)

ENBPBïlDIGTESTS

APR 1987

FULLLOAD-UNIT a

HAY 1988

1967 1988

FIG. 10. TORNESS CONSTRUCTION PROGRAM SHOWING COMPARISON OF ORIGINAL PLAN AND ACHIEVED DATESplant has settled down, trimming of the operating conditionscould permit some further small increase. The expected outturnunit cost is in accordance with the original estimate.The Torness specification incorporates a number of newrequirements. Provision is made for improved access tocomponents and structures within the pressure vessel forinspection and maintenance. Extra provision has also been madefor remote inspection. Figure 11 illustrates these facilities.INSIDE GAS BAFFLE

AREA

INSIDE OFSAS BAFFLECENTRE

INSIDE OFBAFFLEDOME

ANNULUS fiKNUCKLEINSIDE OFGAS BAFFLECYLINDER

CORE

DIAGHID

COMPONENTS

GUIDE.TUBES.DOMECENTREGAS BAFFLEKNUCKLE.UPPERNEUTRONSHIELDRESTRAINTTANK ANDBAFFLEWALL

GRAPHITECHANNELS

DIAGRID.WEBSSKIRT

CONNECTION.SSD PIPES.

ISI

REMOTEVISUAL.(AN ACCESS

TOPERIPHERALTUBESREMOTEVISUAL/

MANACCESSREMOTEVISUAL.LIMITEDMAN ACCESS

AT TOP ENC

REMOTEVISUAL

REMOTEVISUAL/

MANACCESS

REPAIRFEASIBLE

PERIPHERALfHffiiONLY

YES -

NO '

NO „

YES -

INSIDE VAULT

AREA

OUTSIDEOF

REHEATER

MAINBOILERBOILERANNULUSABOVEBOILERGAS SEALBOILERANNULUS

AT BOILERSUPPORTS

SUB BOILERANNULUS

COMPONENTS

SHIELD ANDPERIPHERALGUIDETUBESHANGERS

TAILPIPESHEADERSCASINGSHIELDGAS SEALTAILPIPESPENETRATIONS

TRANS JOINTSCASINGSCASING SHIELDPENETRATIONS

MAIN SEALRESTRAINTSMAIN SEALDIVISIONPLATESSUPPORTS

TAILPIPESCIRCULATORgaîusSKIRT

ISI

REMOTEVISUAL/MANACCESSREMOTEVISUAL/MANACCESSREMOTEVISUAL/MANACCESSREMOTEVISUAL/MANACCESSREMOTEVISUAL,MAN

ACCESSREMOTEWACCESS

REPAIRFEASIBLE

YES

YES

YES

YES

YES

YES

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

20

Experience at Hunterston 'B', after 10 years of operation,justifies the expectation that radiation levels within thevessel will permit man entry on a substantial scale throughoutthe Station's life. Figure 12 shows how the radiation doseacquired by Hunterston 'B' AGR workers compares with typicalvalues in PWR stations.Cracks appearing in some of the welds of the reactor roofinsulation at Hunterston 'B' after ten years' operation havebeen reported in the Technical Press. The thermal insulation ofthe pressure vessel around each fuel standpipe consists of

1.3-,

1.0 -

MAN REM/M.W. (E) Y

USA (PWR.4 LOOPPLANT)

L_

r

L_JI IUJoucncn GERMANSWEDEN PNR

(RINGHALS 2)

SIZEWELL 'B'L _. DESIGN TARGET

HUNTERSTON 'B' AGR

73 74 75 76 77 78 79 1<YEARS

81 82 63 84

FIG 12 COMPARISON OF OCCUPATIONAL DOSE PER UNIT OFENERGY GENERATED FOR DIFFERENT REACTOR TYPES

mineral fibre blankets interleaved with stainless steel foils,the whole being retained by stainless steel tubes and plates.Figure 13 shows the arrangements at Hunterston and the improveddesign at Torness. At Hunterston, certain welds, labelled Aand B in Figure 14, have suffered cracking by thermal fatigue.A small number of these have been removed for inspection andtesting. Research covering structural analysis and flowmodelling is now showing that these welds have been subjected tocyclic temperature variations due to instability of the localgas flow. The solution derived is to fit modified thermal plugs

21

TORNESS

SECONDARY RETENTIONFOR MAIN COVER PLATE

UEL STANOPH>£ INSULATION«FUEL AMCMM.Y NOT 9HOWM)

HUNTERSTON

DIRECTION OFv iew IN FIG 9

SECONDARY RETENTION FOBMAIN COVtR PLATE

FUEL STAWDPIPE INSULATION(FUEL ASSEMBLY NOT SHOWN)

FIG13 ARRANGEMENT OF INSULATION ON LINER ROOF

to the removable fuel assemblies. None of this has safetyconnotations, at least in the short term, because of theredundancy built into the design of the insulation restraintstructure, and normal operations have continued with the consentof the Nuclear Installations Inspector.

SafetyDesign safety guidelines have been adopted by the UK ElectricitySupply Industry which include requirements to meet numericalradiological release targets. The principal target is to aim

22

STANDPIPEINSULATI8H-INNER TUBE

BOTTOMSEALINGRING

MAIN COVER PLATEWELDWELD

FUELSTANDPIPE

CONCRETECOOLINGWATERPIPES

CERAMICFIBREINSULATION ROOFnuur

INSULATION

WELD (ft———WELD (£)CLOSURE RING

WELD

FIG . 14 CROSS SECTION THROUGH BOTTOM OF FUEL STANDPIPE.

for a design which reduces the total probability of anuncontrolled release of activity to the environment to lessthan 10"̂ per reactor year.The design of the Torness safety systems resulting from theapplication of the design safety guidelines is extremely robust,with diverse provisions in the main functions - fault sensing,reactor shutdown and post trip cooling - to maximise reliabilityand afford protection against common mode failures. There aremain and diverse guardlines. The main guardline is a solidstate LADDIC system operating on two out of four logic while thediverse guardline is a two out of three system using relays.The primary shutdown system uses control rods inserted into thereactor core under gravity while, to meet the requirement fordiversity in all safety functions, a secondary shutdown systemis provided which injects nitrogen gas into the core frombeneath the reactor. A manually initiated reactor holddownsystem involving the insertion of boron beads from beneath thecore is also incorporated.Both guardlines will initiate the primary and secondary shutdownsystems. However, only one system is required for reactor

23

shutdown and therefore operation of the secondary shutdownsystem is automatically inhibited once adequate control rods areinserted into the core.

The post trip cooling systems have a high degree of diversityand redundancy. They operate automatically to bring the reactorinto a shutdown mode. The plant is designed so that operatoraction will not be required within 30 minutes of a fault butthe operator can, of course, act to support the operation if asafety system or post trip cooling does not operate correctly.

Due to the high thermal inertia of the core and moderate fuelrating the reactor is very tolerant to failure of the post tripcooling systems. This is illustrated by two examples.First, loss of electrical supplies leads to gas circulator speedrun-down which in turn rapidly reduces gas flow to the reactorcore. The reactor trips from gas circulator under-voltage orunderspeed signals. The diesel generators start from thereactor trip signal and within 50 seconds bring the gascirculators up to their post-trip speed of 15 per cent. Figure15 (lower curve) shows the reactor temperature transient for thecondition in which only the gas circulators and the decay heatboiler in one of the four quadrants are assumed to operate. Itcan be seen that temperatures are well controlled.An investigation of still more severe fault sequences with norestoration of post-trip cooling has shown that, in faults wherethe reactor remains pressurised, it will take many hours beforefuel cladding failure commences and, therefore, correspondinglylonger before any activity release to the environment. Figure15 also shows the long term trends of fuel temperatures in sucha sequence and it can be seen that the rate of fuel temperature

CLAD TEMPERATURE SAFETY LIMIT

FUEL CUDTEMPERATURE - °C

1 2 3 4 5 6 7 8TIME HOURS

FIG. 15 AGR PRESSURISED REACTOR FAULT - LOSS OF FEED

24

rise is modest and that, even after 10 hours, fuel cladtemperatures are a long way from their safety limit of 1,350"C.Secondly, for depressurisation faults, the largest practicablebreak in the pressure boundary is a fracture in the gas coolantclean-up system. A complete severance of the largest pipe inthis system (some 18 cm diameter) would fully dépressurise thereactor in about 40 minutes. The reactor would trip from a lowgas pressure signal and, in the limiting design fault sequence,post-trip cooling is assumed to be provided by only two of thefour quadrants utilising main boilers and gas circulators. Thepost-trip cooling system progressively speeds up the circulatorsfor this fault to compensate for the falling gas density.Figure 16 shows the core temperatures intially risingsignificantly until the reactor trips. This ignores the effectof the auto control system in controlling temperatures at anearlier stage. Even so, the post-trip period temperatures canbe seen to be well controlled.In the same fault the consequences of not initiating any post-trip cooling have been explored. It was found, notsurprisingly, that core temperatures increased more rapidly thanin the equivalent pressurised reactor fault sequence butnonetheless, as Figure 16 shows, the rate of increase intemperature is still modest. The fuel clad (melting) safetylimit would be reached only after about 13 hours. However, withthe reactor pressure lost the fuel pin internal pressures willcause the fuel cladding to creep and some failures would beexpected about five hours after the fault. Even so, this

1100 -i

1000 - FIRST CLADFAILURE

FUEL CLADTEMPERATURE - °C

MINIMUM POST TRIP COOLING

300

TIME - HOURSFIG. 16 A6R DEPRESSURISATION FAULT

25

represents a considerable period of time in which the operatorscould act to restore cooling and, even assuming this conjunctionof failures to remove decay heat, it is inconceivable thatcooling would not be restored in such a timescale.In the AGR, with on-load refuelling, it is necessary to considerthe potential for creating a radiological release as aconsequence of irradiated fuel being dropped during its removalfrom the reactor. The mechanism leading to such an accidentwould be failure of the single tie bar which supports the fuelelements. The nuclear safety implications of such an event havebeen thoroughly investigated and supported by full-scalestringer drop tests. It has been shown that, even at themaximum drop height into the reactor, the automatic tripfacilities provided to deal with this situation would maintainthe damaged fuel clad well below its melt temperature. Theresulting off-site dose to the public would not approach thelevel at which evacuation procedures would be invoked and indeedit would in all probability represent only a small fraction ofthis level.

To summarise briefly then, it has been shown that AGRperformance is excellent under the most extreme design basisfault conditions, but even more reassuring is the time availableto institute remedial measures in those fault sequences, howeverimprobable, in which post-trip cooling systems do not operate.Towards a New AGR

The reactor design for Torness incorporates substantial margins.It is now clear that these margins can be exploited to increasethe heat output from future AGRs.

A contract to produce the design and the appropriate safetyassessment of such a reactor has been placed with the NationalNuclear Corporation (NNC) with finance from the CEGB, SSEB andNNC itself.

The first phase, confirming the main parameters shown in Figure17, has been completed. This was aimed mainly at substantiatingthe higher output but was also necessary to examine any advancein safety criteria emerging from the Industry's practice or theNil's requirements over the years since the last projects werelaunched. In particular, Nil specifically asked for a re-assessment of the gas baffle. This has now been completed andprovides a positive demonstration that the integrity is high,but in addition it is tolerant to large defects.

Work is continuing on the second, more detailed phase of thestudy aimed at the production of a comprehensive design andsafety submission.

Overview

Not surprisingly the Chernobyl disaster brought an immediatefall in support for nuclear power in the U.K. Despiteexplanations of the differences in design and operation of U.K.reactors and differences in licensing procedures, the political

26

PARAMETER

No. OF FUEL CHANNELS.REACTOR HEATMEAN CHANNEL RATINGPEAK CHANNEL RATINGGAS PRESSURE UNDER TOP SLABCIRCULATOR GAS OUTLET TEMP.BULK CHANNEL GAS OUTLET TEMPCIRCULATOR POWER CONSUMPTIONBOILER STEAM FLOWSTEAM PRESSURE AT H.P.TURBINESTEAM TEMPERATURE AT H.P.TURBINEFEED TEMPERATUREGROSS GENERATED POWERNET POWER SENT OUT

UNITS

MWMWMW

bar a°C°C

MW(e)kg/sbar a°C°C

MW(e)MW(e)

HUNTERSTON B(TYPICAL 1986)

30815354.986.4439.32B663440.5515155478156640592

TORNESS

33216234.896.2641.029863942.152S167538158705645

PROPOSEDA.G.R.33217405.246.7042.529763945.2552160538142763700

FIG.17 A G R PARAMETERS

parties in oppostion are calling for either phasing out ofnuclear power or a period of re-assessment before authorisingnew construction.The report from the Inspector of the Sizewell Inquiry into theproposal by CEGB to build a PWR is expected in the immediatefuture. Much now depends on the content of this report and howthe Government, still strongly supporting nuclear power, decidesto proceed. With the likely size of nuclear ordering programme,it would be unrealistic to build more than one nuclear system inthe U.K. Even with the most optimistic view of the nuclearordering programme, this would incur heavy penalties in costsand performance in maintaining the infrastructure.

Clearly there would be no incentive to consider a change in theU.K. from well-tried CÛ2 cooled technology to embark on thelengthy business of bringing any new gas-cooled system such asthe HTR to commercial success. Notwithstanding the excellentsafety characteristics of the AGR which have permitted sitingclose to urban locations, there has as yet been no seriousinterest in combining these with heat production forlarge industrial processes. Nevertheless the continuance ofgas-cooled technology in its present form in the UK must be ofcontinuing assistance in the efforts to develop HTGR in Europeand the US, in the better appreciation of what gas-cooledtechnology has to offer, including the transfer of engineeringexperience and development know-how, much of which can bedirectly related to HTGR requirements.

27

HIGH TEMPERATURE GAS-COOLED REACTORS

R. SCHULTENInstitut für Reaktorentwicklung,Kernforschungsanlage Mich GmbH,Jülich, Federal Republic of Germany

Abstract

The historical development in energy technologyhas the goal of realizing high temperatures and highefficiencies for the energy conversion processes. Thehigh temperature reactor (HTR) realizes this goal inthe best way compared to other reactor types; it usesceramic materials and helium as the heat transfermedia. The development of the fuel elements on thebasis of coated particles is considered to be finished.Future applications are the production of electricityand the production of process heat for other areasof the energy market. Originally the thorium/uraniumfuel cycle has been envisaged, today the low enricheduranium fuel cycle is of interest, because of thelow prices of uranium. Typical for the HTR is itsflexibility with respect to different concepts. On thebasis of the existing prototypes AVR, THTR-300 andForth St.Vrain the following two development lines arein discussion: reactors of small unit size and reactorsof medium unit size. The safety qualities have beendemonstrated by the operation of the AVR reactor inJülich, which is now in operation for 19 years. Theseexperiences will be of great importance for futureplants.

1. Fundamental conceptsThe historical goal of the technology of power

station, as well as in energy technology in general,is the realization and application of high temperaturesfor the processes of energy conversion. The high tem-perature reactor realizes this goal to the best compared

29

to other reactor 'types. With high temperatures in therange between 700 and 1000 °C high efficiencies arerealized for the production of electricity and otherapplications in the energy market are possible. Therealization of high temperatures is possible becauseof the application of ceramic materials. In combina-tion with requirements from the neutronic side thematerial for the moderator and for the constructionof the HTR is graphite. As the coolant the inert gashelium is applied. The pressure of the primary circuitlies between 40 and 60 bars. On the basis of theseprinciples research, development and demonstrationwork has been done since about 25 years in the FederalRepublic of Germany, in the United States, in theUSSR, Great Britain and in Japan as well as in someother countries.

One of the most important successes of the R+D-work in HTR technology is the technique of the coatedparticles, FIG. 1. in the Federal Republic of Germanythe pebble bed typ fuel element has been developedand is used. The pebble contains the coated particlesin an inner part, which is surrounded by an outerlayer free of coated particles. This concept of afuel element is considered to be an industrial product.

Carùon layers

HTR fuel element

FIG. 1 : HTR fuel element

30

Its development is also a success. An important fea-ture is that very high burn-ups can be achieved. Therelease of fission products during normal operationis very small. The temperature stability of the fuelelements is also in the range of 1600 to 1800 °C verygood. By experiments it has been shown that the releaseof fission products in this temperature range is mi-nimal. Because of this good temperature stabilitythe concept of the HTR-Modul-reactor and the conceptof the HTR-industrial-reactor have realized a par-ticular safety concept.In this concept practically no radioactivity is re-leased to the surroundings in the heaviest accident.In addition to that this concept realizes repairabilityafter the heaviest accident.

In the Federal Republic of Germany the pebbletype fuel element has been developed as well as inthe Soviet Union. In contrary to that the bloc typefuel element has been developed in the USA, FIG. 2.The main feature of this fuel element is a hexagonalbloc made from graphite. This bloc contains 'bore-holes

Blockförmiges Brennelement

1 Kühlkanal2 Brennstoffstab3 Abbrennbares Neutronengift4 Bohrung für Steuerstab5 Greiferbohrung6 Bohrung für Absorberkugeln

FIG. 2: Prismatic fuel element

31

for the fuel sticks and bore-holes to guide the coolanthelium. The bloc type fuel element is applicable forsmaller and larger reactor cores.

2. ApplicationsThe HTR is first of all, as well as other reactor

types too, developed for the production of electricity.The presently established technology is, to producesteam and to use that steam in turbines for the pro-duction of mechanical and electrical energy. The HTRrealizes conditions for that steam as they are usedin conventional power plants. Fundamentally the appli-cation of gas turbines is also possible in directand indirect loops. This technology has been inten-sively developed in the seventieths. During this timethe further realization of gas turbines could not becontinued because the know how on the behaviour ofthe primary loop was not sufficient. The advantagesof the application of gas turbines are remarkable,therefore its application may be of interest in thenext decades.

Beside the production of electricity the HTR isalso applicable for the production of process heat.In cogeneration of electricity and process steam abroad variety of applications does exsist in chemicalindustry. An area of particular interest in thecoming decades is the production of oil via enhancedoil recovery, here injection steam can be producedwith the HTR. An intensive research, development anddemonstration program has been done for the applicationof nuclear process heat for the refinement and conver-sion of coal, oil and gas for the production of en-vironmentally benine liquid energy carriers, which canbe used as fuel for motor cars and for heating purpo-ses. The advantages are the following: applicationof nuclear energy instead of fossil fuels for theprocess heat, reduction of problems with respect tothe environment, particularly the reduction of the

32

Problems of carbon dioxide. Today the latter applica-tions are not possible because of the low prices ofoil. But it is expected that these applications willbe economically attractive in the coming decades andit is expected that they will be of great importancein future because liquid energy carriers are requiredin the energy world supply and because the C02~effectson the climate may become a problem.

3. Fuel cycleThe fuel cycle of the HTR is very flexible. The

original goal of the development work in the worldhas been the application of the thorium/uranium-fuelcycle. In calculations it have been shown that withthis fuel cycle a near breeder can be realized. Inthis conference there will be given a report, whichshows, that even a breeding process can be realized.Our own calculations, which have been finished fiveyears ago, have shown that a conversion factor ofabout 0.8 is economically the optimum, this is ofcourse influenced by the costs of the reprocessing ofthe fuel. A conversion factor of about 0.8 realizesa utilization of the nuclear fuel which is five timeshigher compared to a fuel cycle without reprocessing.

Today the fuel cycle with low enriched uranium(8 %) without reprocessing and with direct disposalis prefered. The reason is the low price of uraniumand its sufficient availability. This fuel cycle hasthe advantage that almost all plutonium produced inthe reactor is used. In addition to that the producedplutonium contains much higher isotopes; therefore itis not dangerous with respect to proliferation. Theutilization of uranium with this fuel cycle is almostas high as in the fuel cycle of a .light water reactor(LWR) with one reprocessing and recycling of the pro-duced plutonium. In contrary to this fuel cycle pro-posed in the Federal Republic of Germany,, a fuelcycle with an enrichment of 20 % will be used in theunited States.

33

4. HTR conceptual designsThe pebble type fuel element allows a high flexi-

bility with respect to the dimensions of the reactorcore. It is possible to realize reactor cores withsmall diameters and relatively large heights, as wellas cores with a large diameter and smaller heights.The principle of the pebble bed reactor is explainedin FIG. 3.

top reflectorgraphite guidefor shutdownrods

linerthermalshield

reflectorbottom

supportstructure

top shield

biological— shield

— side reflector— reactor core

_ outer vessel— inner vessel

fuel elementdischarge pipe

FIG. 3: The principle of the pebble bed reactor

The pebble bed is cooled by the heat transfermedium helium. The pebble bed type fuel elements aretransported by a pneumatic system to the upper partof the core, and they are extracted from the lowerpart. In principle two fuel loading modes are possible:there is the once through than out mode and the multiplemode, in which the fuel elements are recirculated.Both fuel loading modes have been investigated inten-sively. During this conference the differences ofthe two modes will be discussed.

An additional flexibility is realized becausetwo types of reactor pressure vessels are feasible:

34

the prestressed concrete vessel and the steel vessel.The prototype for HTRs with steel vessel is the AVRin Julien. This reactor has during its almost 20 yearsof operation resulted in fundamental know-how on theHTR, FIG. 4.

/ Sltam Header2 Containment3 Outer Pressure Vessel4 Inner Pressure Vessel5 Steam Generator6 Quench Tank7 Upper Reflector0 First Biological Shield9 Pebble bed Core

10 Lateral and Bottom Reflector11 Fuel Oischarge Pipe12 Gas Purification System13 Dismantling Machinery14 Second Biological ShieldÎ5 Charge room for Fuel ElementsK Cooling Gas Circulators17 Dismantling Machinery for Circulators18 Components of Fuel-handling System19 Material Lock20 Heavy Load Elevator21 Circular Bottom Support22 Storage Channel for Spent

Fuel Elements

FIG. 4:

AVR-reactor in Jiilich

The German reactor constructors have on the basisof the AVR developed interesting concepts for smallHTR units. At first I would like to mention the Modul-HTR of the company Interatom, FIG. 5. In additionto that I would like to mention the HTR-100-conceptof the company HRB, FIG. 6. Both proposals are presented

35

FIG. 5: Modul-reactor (INTERATOM)

+ 42,0

+ 12,5

1 Spent fuel building2 Reactor auxiliary building3 Reactor building4 Residual heat removal building5 Turbine hall6 Industrial steam plant

FIG. 6: HTR-100-industrial reactor (HRB)

in this conference. It is my opinion that these smallHTR units are of interest for all applications inwhich smaller power rates are of interest. By thecombination of several small units power plants in therange of 200 to 500 MWel may be realized. In the USAtoo the idea of smaller reactors is discussed. Thisinteresting development will also be presented inthis conference.

36

The second concept of the HTR is characterized bythe prestressed concrete vessel. A prototype of thiskind is the THTR-300 in Schmehausen. This reactorwent into operation during this year and has reached100 % electricity production recently, FIG. 7. Thisreactor is the base for the next HTR, the HTR-500.The planning of the HTR, which is done by HRB and a

FIG. 7: 300 MWe nuclear power plant Hamm-Uentrop

group of utilities has started recently. On this con-ference there will be report about it. Parallel tothis there has to be mentioned the Fort StVrain-reactorin the USA. This reactor is in operation since years,about the result a report will be given during thisconference.

5. Safety qualitiesThe AVR-reactor in Julien is excellently suitable

to demonstrate various safety qualities of the HTR.In particular it has been proofen very good as a plantfor the tests during the development for fuel elements.

37

The AVR operates at a mean helium outlet temperatureof 950 °C. Even at this high temperature the primarycoolant is relatively clean. The contamination isvery low. In the AVR-reactor experiments have beenperformed in which the loss of the coolant withoutscram has been simulated. The results have been thatthe negative temperature coefficient of the reactivityis fully effective and that a stabilization of thetemperature niveaus in the reactor core takes place.This is due to inherent transport mechanisms for heatin the core. In 1978 there happend an incident byan ingress of water from the steam generator. Thisincident has produced much know how on effects onsteam and water to the reactor core. The operationof the AVR will continue in the years 1987 and 1988.The main purposes are experiments which are safetyrelated.

The concept of the small HTRs in the range of200 to 250 MWth has been developed on the basis of theexperiences gained in the AVR. The most important as-pect of the concept of the small HTR is that in heavyaccidents the maximum temperature of the core doesnot exceed 1600 to 1800 °C in the reactor core. Becauseof this the release of fission products from the fuelelements is rather small. Therefore these reactorshave a particular safety quality. In addition to thatthere is a fundamental possibility of repairabilityafter heavy accidents. It is my opinion that this reac-tor concept because of its particular safety qualityis of great importance for the future.

As mentioned before there are the plans from theutilities and from the company HRB to realize a follower -reactor to the THTR with a power of 500 to 550 MWe.In this concept particular safety qualities of theconcrete vessel will be used. Right now an interestingexperimental work is done in the KFA Jiilich on thebehaviour of concrete vessels. The goal is to demon-strate heat resistance of the concrete upto tempera-ture of 1300 °C.

38

6. Status of the developmentThe status of the research, development and demon-

stration work on the HTR is summarized as follows:- the development of the pebble type fuel element andthe bloc type fuel element is almost finished. Thehigh quality of the fuel elements hase been demon-strated by the operation of plants as well as bylarge experimental programs.

- There have been developed concepts for small HTR onthe bais of the operational experiences of the AVR-reactor. In the Federal Republic of Germany as well inin the United States there are undertaken effortsfor the realization of such small reactors and forthe demonstration of their particular safety quali-ties. The planned small reactors in the USSR alsobelongs to this category.

- And finally with the successful operation with theTHTR-300 it might be possible to offer of follower-reactor with medium unit size.

39

EXPERIENCE WITH GAS-COOLED REACTORS(Session B)

DEVELOPMENT OF A HIGH OUTPUT AGR

B.A. KEENNational Nuclear Corporation Ltd,Knutsford, United Kingdom

Abstract

The paper outlines the design of a new AGRbased on the Heysham II/Torness reactors which arenearing completion of their construction stage. Theintent is to replicate as far as possible the HeyshamII/Torness design in meeting the target stationgenerator output of 1526 MW(e), and this can beachieved by design changes only to the HP steampipework.

Two additional design options, a ventedcontainment and a dry refuelling route, are alsounder consideration.

1. INTRODUCTIONThe Advanced Gas-Cooled Reactor Power Stations

now being constructed at Heysham in Lancashire,England, and Torness in East Lothian, Scotland,represent the current stage of development of thecommercial AGR. Each power station has two reactorturbo-generator units designed for a total stationoutput of 2 x 660 MW(e) gross, although it iscurrently intended to uprate this as far as possibletowards the figures discussed later in this paper.

The construction of these new AGRs has been toprogramme and within budget. Fuel loading for thefirst reactor at Heysham is expected late in 1986with the other three reactors following over thesubsequent twelve months. The design of bothstations has been based on the successful operatingAGRs at Hinkley Point and Hunterston which have nowbeen in service for almost ten years, although minorchanges were made to meet new safety requirements andto make improvements suggested by operatingexperience.

At the Sizewell 'B1 Public Inquiry, the CEGBgave a commitment to retain the AGR option. NNC isengaged in the definition of the improved AGRdesign. In order to use the invaluable experiencegained since the last two AGR stations were ordered,it was decided that the new designs should bedeveloped from the Heysham II/Torness design. Aspart of this study, it was necessary to address the

43

specific safety requirements for future AGRspresented to the PWR Inquiry by the NuclearInstallations Inspectorate.

In the interest of continued safety improvementaimed at mitigating the consequences of accidentshowever unlikely these may be, consideration is alsobeing given to a design of vented containment. Themove by the CEGB to construct a national dry fuelstore has created interest in a dry refuelling routeas another possible development. It should be notedthat the design study is not yet complete and nodecisions have yet been taken on these options.

2. REACTOR DESIGNThe new design of power station has two

identical reactors of 1740 MW thermal output drivingindividual turbo-generators giving a station grosselectrical output of 2 x 763 MW(e). The additionalrating of the turbing/generators will be accommodatedby small changes to the design of the HeyshamII/Torness units. Principal dimensions and designparameters are given in Table 1.

The reactor is contained in a single cavityprestressed concrete pressure vessel (PCPV), asection through the vessel being shown in Figure 1.The reactor core is supported on a diagrid andenveloped by the gas baffle, which is welded to thepressure vessel liner floor to prevent movement inthe event of an earthquake. The gas baffle enablesthe graphite moderator and other core structuralcomponents to be adequately cooled during normaloperation.

The reactor fuel is made from slightly enricheduranium in the form of sintered uranium oxidepellets. The pellets are assembled in multi-startspirally ribbed stainless steel cans to form fuelpins, 36 of which are enclosed in a graphite sleeveto form a fuel element. The fuel elements are to thenew Stage II design with single thick sleeves toimprove their resistance to fuel handling loadings,and with a lower flow resistance. Each fuel assemblycomprises a bottom support unit and reflector, eightfuel elements, a top reflector and a lower gag unitthreaded onto a tie-bar which is suspended from aplug unit which incorporates a flow control gag unit.

Eight gas circulators deliver carbon dioxidegas at 42.5 bar a and 297°C into the plenum below thediagrid. Heat is transferred to the gas from thefuel raising its temperature to 645°C. Main boilerunits are arranged circumferentially in the annulusformed by the gas baffle and the reactor pressurevessel liner. Each boiler has high pressure and

44

TABLE IPRINCIPAL DIMENSIONS AND PARAMETERS

Pressure VesselInternal heightInternal diameterWall thicknessTop slab centre thicknessBottom slab centre thickness

21.9m20.3m.8m5.5.4m7.5m

Reactor Primary CircuitGas baffle internal diameterGas baffle dome height above liner floorCore restraint tank outside diameterType of main boiler

Number of main boilersNumber of boiler units per main boilerType of decay heat boiler

Number of gas circulatorsType of gas circulators

Reactor CoreNumber of fuel assembliesNumber of control assembliesNumber of secondary shutdown system channelsHeight of core (including neutronshields)Overall width of core across flats

13.9m19.9m13.6m2 start, oncethroughSerpentineplatenrectangular unit431 start, oncethroughSerpentine platen8 (2/quadrant)Centrifugal,single stageencapsulated,varaible inletguide vanes

3328916312.8m12.4m (regular16 sided polygon)

Reactor ParametersReactor heatBulk circulator outlet gas temperatureBulk fuel channel outlet gas

Net circulator flowGas pressure at circulator outletPressure difference across gas baffle domeTotal primary circuit pressure dropPeak channel power

1740 MW297°C645°C at top offuel temperaturestack4697 kg/s45.2 bar a2.56 bar3.02 bar6.7 MW

Boiler ParametersFeedwater flowFeedwater temperatureHP steam temperatureHP steam pressureReheater steam temperatureReheater steam pressure

552 kg/s142°C538°C160 bar a539°C40.7 bar

Turbine - Generator ParametersTurbine heat rateNominal gross electrical output/reactorNominal net electrical output for stationDesign lifetime

2.31 kJ/kJ(e)763 MW1400 MW30 full poweryears

45

TYPICAL FUCl SWHOPIPE —

TOP MAN ACCESS PENETRATION

—COHTHOL SIANOPIPE

•PRESSURE VESSEL

BOIL £ R REHEAT MLET-^ AND OUTLET

PEKEmitON*

-REHEAT CASINO

SECONDARY SMUIDOW« ROOK

FIG 1 REACTOR PRIMARY CIRCUIT SECTIONAL ARRANGEMENT OFREACTOR UNIT BASIC VAULT

reheat sections with separate feed and steampenetrations through the outer cylindrical surface ofthe pressure vessel. The main boilers and gascirculators are physically divided into four separatequadrants by division plates at the circulator inletplenum below the boiler seal. During operation, feedwater is supplied to the boilers at 142°C and 217 bara giving steam temperatures of 538°C. The gascirculators are driven directly from grid suppliesusing 45 MW from the station output. Gas flowcontrol is achieved by adjusting the angle of guidevanes at the inlets of each circulator. A totallydiverse secondary system for the removal of decayheat is provided via the decay heat boilers (DHB).These are commissioned automatically following a tripand have their own feed system and heat sink.

The primary system for control and shutdown ofthe reactor comprises absorber rods and drives housedin standpipes in the top cap of the PCPV. Theactuator and chain store is mounted between a closureunit and radiological shield plug through which thechain passes and from which hang the articulatedtubular control rods.

46

A secondary shutdown system (SSD) designed toinject nitrogen into the core is provided as adiverse means of shutting the reactor down in theremote fault situation, should a predetermined numberof control rods fail to enter into the core whenrequired. This system operates from the SSD roombeneath the PCPV. Boron glass beads can be injectedby the operator into the core to provide a long-termreactivity hold-down capability in the event ofcontinued failure to release the primary shutdownrods before the reactor is depressurised.

Both reactors are served by the fuel handlingfacilities located between them, a single fuellingmachine and other shared equipment. These facilitiesprovide for the storage and assembly of new fuel, thetransfer of fuel in and out of the reactors and decaystorage tubes, and subsequent dismantling of thefuel, servicing of reusable plug unit components anddisposal of debris, the subsequent pond storage ofirradiated fuel and finally its despatch from site.3. DESIGN CONSTRAINTS

The main constraint on the development of thedesign was to replicate as closely as possible theHeysham/Torness design. Only changes to meet newsafety requirements or simplification of the designwhich would not lead to significant alterations tostation layout were considered. Operatingconstraints were also applied to ensure that theexperience of the existing AGRs were still valid andthat the safety case prepared for Heysham/Tornessremained applicable.

Within these constraints, and with theexceptions given below, it has been shown thatsufficient flexibility remains in the designparameters to allow for uncertainties and to ensurethat the target station generator output of 1526MW(e) will be achieved for the operating lifetime ofthe plant of the equivalent of 30 full power years.It is possible that greater output might be achievedafter commissioning has demonstrated the expectedavailable margins.

4 HAZARD DESIGN4.1 Earthquakes

A new enveloping Safe Shutdown Earthquake (SSE)has been defined for all future nuclear stations inthe UK. This SSE is based on historical evidence ofseismic activity on the British mainland and has apeak acceleration of 0.25g. The effect of thisearthquake on the buildings and plant will bedependant on the soil structure at the site. The SSEloadings have been derived for the two extremes of UK

47

site, with the most onerous conditions being used asthe design basis. All essential plant and structuresare designed to survive this design basis SSE withoutpreventing the shutdown and post-trip cooling of theplant. In order to prevent a sudden loss ofcapability at earthquakes with accelerations justbeyond the design basis, the aim of the design hasbeen to provide a conservative response at the SSEsuch that adequate shutdown and cooling would beavailable for earthquakes up to 0.35g and beyond.4.2 Aircraft crash

The frequency of aircraft crash on thevulnerable areas of the station has been assessed fortypical sites "outside" and "within" Areas of IntenseAerial Activity. The study confirmed the impactfrequency was very low for both areas and that withthe high degree of separation and segregation ofplant in the Heysham II/Torness layout, the risk ofcausing a radiological hazard was not significant.The addition of a fire fighting system in the reactorpile cap area to contain aviation fuel fires reducesthe risk further.

5. DESIGN CHANGESThe effect of the higher output on the design

of Heysham and Torness has necessitated significantdesign changes only to the HP steam pipework toaccommodate the higher steam flows without excessivepressure drops, and to meet the extended stationlifetime.

Manufacturing and construction experiencegained during the Heysham and Torness projects hasbeen used to make detailed changes in various areasof the design to simplify manufacture, easeconstruction and/or to reduce costs. None of thechanges adopted will significantly affect the objectof replication.

The safety case for on-load refuelling atHeysham II and Torness has only been presented forrefuelling in batches as part power levels withconsequential loss of output. Simplification of theprotection systems will enable the time taken foreach batch to be reduced, while further generic workon refuelling is expected to provide the means forincreasing the present limit on refuelling powertowards full power and single channel refuelling.These improvements will give a significant increasein reactor availability.

48

6. VENTED CONTAINMENTAGR and Magnox Safety Cases have always been

made on the basis that no credible accident can beidentified which would require secondary containmentof the reactor. Until recently, for HeyshamII/Torness and earlier AGRs, the probability of adropped fuel assembly during refuelling had beenconsidered to be so low that it was not necessary totake account of it in the safety cases. However,should it occur, vented containment of the sub-pilecap volume would reduce the activity release fromsuch an accident.

Other potential release areas in the reactorbuilding have also been considered, even though noaccidents have been postulated which would requiresuch provision. The key areas identified for C02discharges are:

Pile CapQuadrantsSSD RoomC02 Bypass and Clean-up CircuitFuel Decay StoreEach of the five areas will have its own high

integrity ventilation system, broadly based uponthose which currently exist on Heysham II/TornessAGRs.

In the event of a breach of the pressurecircuit, the normally running ventilation plant isisolated and a hot gas release route is opened todischarge the escaping gas directly to atmosphere.

The new plant connected to the aboveventilation routes by normally closed dampers, wouldcomprise heat exchanger, high efficiency particulateair and charcoal filters, and extract fansdischarging to a new stack. The plant would start upautomatically on high temperature or C02concentration and when the discharge flow through thehot gas route reduces to the capacity of the newplant, dampers would close to route all the dischargethrough the new plant. For discharges within thecapacity of the filter system, isolation would beimmediate.

In the unlikely event of a dropped assembly,fuel damage and activity release would be immediate,therefore the new plant would considerably reduce theconsequences. Other breaches within the designbasis, although initially exceeding the capacity ofthe plant, would not give rise to early fuelfailures. For faults beyond the design basis, theplant would mitigate the consequences by reducing thetotal release of activity.

49

7. DRY FUEL ROUTERecently the CEGB have decided to build a large

central dry store away from the reactors, capable ofaccepting irradiated fuel and acting as a bufferbetween them and the final reprocessing atSellafield. This new development makes it logical toreview the possibility of adopting a dry fuel routefor any future AGR.

The present sequence at Heysham/Torness is forirradiated fuel to be placed in one of 30 bufferstorage tubes after removal from the reactor by thefuelling machine. The buffer storage tube is watercooled, the heat being transferred from the fuelstringer to the water by natural circulation ofpressurised CC>2. After 28 days the decay heat hasreached a level where cooling by air at atmosphericpressure is adequate and the fuel is transferred toone of two Irradiated Fuel Disposal Cells to bedismantled. The elements are then transferred to thepond where it is stored for about 100 days beforedispatch.

The dry route would follow the same sequence,but the number of buffer storage tubes would need tobe increased to 50 to allow storage for 100 days.The increased decay time is required to permit thedecay heat to reduce to acceptable levels before thefuel is placed in the dry store after passing throughthe IFD cell. In the dry store, the fuel will remainfor a minimum of 265 days before it can be despatchedfrom site in a dry fuel flask.

The dry store is separate from the reactorbuilding with loading and unloading cells at one endso that the store may be extended during station lifeshould this become necessary. The initial capacityproposed is 1280 elements, stacked 16 high in 80tubes, cooled externally by natural circulation ofair.

There are no technical obstacles and a dryroute would remove any doubts about the long-termstorage of AGR fuel after passing through a pondsystem, although development of a dry road transportflask would be required.

8 CONSTRUCTION PROGRAMMEBased on the experience gained at Heysham and

Torness the construction period from start ofpermanent works to commercial load of the firstreactor would be reduced by 3 months to 75 months.

50

9 LICENSING PROGRESSAs the high output AGR is very similar to

Heysham and Torness, their Station Safety Report,SSR, forms the basis for the licensing of any futureAGR in the UK. The SSR was submitted to the Nil in1986 in preparation for start-up of the first reactorat Heysham. An assessment of the extra requirementsraised by the Nil at the Sizewell Inquiry has shownthat while extra safety studies will be required, nosignificant design changes are expected to be made.

10 CONCLUSIONSThe successful progress of the Heysham and

Torness projects in the UK coupled with the goodoperating experience of the similar Hinkley Point andHunterston reactors forms a strong base for thedesign of a high output AGR. In producing the newdesign, the major aim has been to replicate theHeysham and Torness AGRs with the minimum ofchanges. The increase in output of 15% to 1526MW(e) gross generator output coupled with a reductionin the construction programme and costs improves thecompetitiveness of the new design.

With a modest investment, two furtherdevelopments, vented containment and a dry refuellingroute, could be added. The former would mitigate therelease from faults which are beyond the presentdesign basis. The latter would be a logical step inview of the intention to build a national dry fuelstore.

An assessment of the Nil requirements for afuture AGR in the UK as stated at the Sizewell PublicInquiry has shown that no major difficulties areexpected in obtaining an operating license, althoughdesign substantiation will be required.

51

TECHNICAL EVOLUTION AND OPERATION OFFRENCH CO2 COOLED REACTORS (UNGG)

Y. BERTfflONCEA, Centre d'études nucléaires de Saclay,Gif-sur-Yvette, France

Abstract

The technical evolution of the five French C0£ cooledreactors (UNGG) from 1981 to 1986 needs to be outlined. Thesetechnical evolutions concerned the fuel element of Bugey 1 whichis now slightly enriched, as well as the load reduction operationrequired by the grid. In addition work in underway to increasethe safety at the two St Laurent units, or to repair the hotsteel upper-structures of Chinon-3 unit.

1. INTRODUCTIONThe technical evolution of the five French C02 cooled

reactors (UNGG) from 1981 to 1986 needs to be outlined. Thisevolution will be illustrated by examples taken from a surveymade at each nuclear power plant. I will speak of the steelcorrosion and the repairs (ISIS) at the Chinon-3 unit, theerosion - corrosion of heat exchanger and safety at the twoSt Laurent units, thé graphite corrosion and the load-reductionoperation at Bugey.

Evolution in the last years is illustrated by figure 1 whichgives the cumulated load factors of each unit and the predictedcumulated load factor quoted PEON. One can see the shutdown forrepair.at Chinon-3 unit started on May 4, 1984. But figure 1 does

i — l — i — l — l l — i — i — i — l — \ — i — i — i — i1 2 34 56 7 8 9 10 11 12 13 14 15 16 17 18 18 20

nb.d'annee

FIGURE 1 : UNGG cumulated load factors versus time

53

not give a correct idea of the availability factor of the lasttwo years ; from now on, the predominance of nuclear generationin the French electrical network will require the nuclear unitsto take part in load-following and frequency adjustment of thegrid. In France the PWR units equiped with the control mode Gallowing power variations are cheaper than UNGG units. Also thegrid requires load reduction of the UNGG reactors.

2. THE CHINON NUCLEAR POWER PLANTThree UNGG nuclear reactors are located in Chinon.. The CHINON-A1 reactor which shutdown definitively in 1973

now serves as a museum and this since February 3, 1986.. The CHINON-A2 reactor was shutdown for economical reasons

in June 1985 after 20 years in operation. Its cumulative loadfactor was 73.7 % i.e. close to the PEON prediction (figure 1),and have a last run of 869 days without scram. The management ofthe fuel load was optimized and during the last 8 months no newfuel element was loaded. The radial shuffling of fuel elementsand the overmoderation obtained with the feeding of the fuelchanel by graphite log provided the needed reactivity to operateand produced 970 GWeh of extra electricity. The average availabi-lity factor of the last five years was 86 '% with a maximum of94.5 % in 1984. During the life of this reactor, 133,374 fuelelements were loaded with only 6 clad failures. Some shuffledfuel elements are now in hot cells for examination. Good resultsare expected and the radial shuffling will be planned for theend-of-life of other UNGG reactors in order to save fuel. Table Ishows the reliability of UNGG fuel elements.

TABLE I. RELIABILITY OF FUEL ELEMENTS (30/06/86)

Number of

. unloadedfuel

elements. loaded

fuelelements

. cladfailures

failure rate

Graphite Core fuel elementsSICRAL F1

369,020"

542$250

10

< 2 10~5

Annular fuel elementSICRAL F1

75,550

88,140

3

< 4 10~5

. Thé CHINON-3 reactor has been shutdown since May 4, 1984in order to repair the steel corroded upper structure of thereactor. Figure 2 shows a vertical section of this unit. Thecorrosion did not have the same consequences on the Chinon-A2reactor whose structures were bolted instead of welded (figure 3),nor on the subsequent reactors because the CÛ2 steel corrosionwas already known. This operation of structure repairing is cal-led ISIS and the cost up to mid-1987 will amount to 300 MF. The

54

Tbntde manuhpnhon15 r.

feiton

FIGURE 2 : CHINON-3 reactor vertical section

CHINON 2(20 years)

CHINON 3

Oxide formation

Bolted

Hot screak temperature ;

>. FAILURE WITHIN 10 YEARS

Welded

FIGURE 3 : Mild steels corrosion mechanisms by C02

erection of a building was necessary to shelter a true-scalemodel of a sixth of the upper structure. Rigorouscorrespondence between the model and the upper structure isobtained by a laser telemetry done on the reactor. Five robotsare necessary : three on the reactor, i.e., one for the camera,one for the tools and one for the metal to be welded. The twoothers are on the model for learning purposes. The reactor willoperate again in mid-1987. Steel corrosion was responsible for

55

the nominal power lowering : 360 MWe instead of 480 MWe (seeTable II). On the other reactors, the nominal power was reducedfor different reasons. But local defects where steel corrosionis involved have been noted and their evolution is being careful-ly watched.

TABLE II. POWER OF UNGG REACTOR

UNIT

CHINON A-2CHINON A-3

SL1SL2

Bugey

Design Power(MWe)

480480515540

Effective Power(MWe)

360390515540

Difficulty

Stell corrosion[ erosion-corrosionon heat exchangergraphite corrosion

3. THE BUGEY NUCLEAR POWER PLANTA UNGG unit is located on the Bugey nuclear power plant.

This unit differs from the others in its fuel elements which areexternally and internally cooled, as well as in the C02 pressurewhich is 43 bars instead of 29 bars. These two factors allow fora greater specific power. But high specific power and high pres-sure mean high rate of graphite corrosion. The maximum local gra-phite corrosion measured is 22.4 % at 8.58 aepp (year equivalentfull power). The mechanical behavior of such a corroded block ispoor, and mechanical and seismic calculations are under way inorder to evaluate a maximum acceptable level of corrosion and todetermine the remaining lifetime of the reactor. To day the life-time is estimated to be 2 aepp. So this unit is now considered asan energy tank. In France the marginal cost of PWR is the chea-pest and these reactors operate in load following mode. So froman economical point of view the grid necessitates a load reduc-tion of these UNGG reactors. Therefore, the optimal managementof this energy tank prompts the operator to define several ope-ration levels acceptable by the fuel and the reactor (Table III).

TABLE III. OPERATION LEVEL

Power (MWe) Comment540470

400, 300, 200

Exceptional grid demandNo more than 500 hours/yearNormal operation level upon grid demandTen times a year.

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This operating mode requires more from the fuel elements,so in order to increase the load-reduction number the fuel beha-vior is examined in hot cells. To reduce the graphite corrosion,methane is injected into the CÛ2 with a 450 VPM level. The decom-position of methane produces a carboxyhydrogenated deposit whichreduces the heat transfer from the fuel, and the reactivity withthe neutron capture on hydrogen. The high hydrogen content in thestack has made necessary the use of a slightly enriched uranium(0.76 % instead of 0.72 %) since June 28, 1984. This enrichedfuel gives better power distribution, and flexibility. The lowincrease in cost is compensated for by saving fuel in the outerpart of the reactor. To increase the burnup of the outer fuelradial shuffling is not used at the moment.

4. THE SAINT-LAURENT NUCLEAR POWER PLANTTwo UNGG units called SL1 and SL2 are located on this site.

Figure 4 shows a vertical section of the SL1 reactor. The SL2 unitis very similar, as is the Vandellos unit in Spain. The nominalpower was reduced (Table II) because of a erosion-corrosion problemof the heat exchanger. This phenomenon originated from a concep-tion flaw, and the vaporization level was changed in reducing thepower. Meanwhile, during the first operating years, the exchangerwas affected by erosion-corrosion, and today some leaks still occur.

The heat exchangers are divided into two separate parts andthe leaks affected mainly one of the two halves in both SL1 andVandellos. So, some other phenomenon may also be involved. Thewater quality is important, and at Vandellos better results wererecently obtained with AMP than with morpholine.

Over the last five years, however, work has been focussed ona réexaminâtion of the Safety Report in the light of feedbackexperience. Among these works are :

. EAR use as ultimate emergency system.

. The building of an indépendant emergency control roomnamed BUS.

. Increasing blowing availability by reducing the common-causes failure with fire retarding bulkheads and the separationof cableway, and by supplying auxiliary sets with main steam.

. The possibility, in case of scram, of feeding turboblowerwith low quality main steam. This forced running provides morethan half an hour of blowing.

. The use of false tulip-shaped fittings on each fu.el chan-nel in order to prevent plugging.

In addition to the erosion-corrosion problem and work onsafety, a third point concerning the two Saint-Laurent units isa thirty-year extension of their life. To "achieve this extension,the consequences of steel and graphite corrosion will have to bestudied.

* heat exchanger cooling the outage unit.

57

ENSEMBLE DU RÉACTEURCoupe verticale

28.500 J32.300^131.100

138 câbles

92.250

Nota : la (urbo-soufflante est ramenée dans le plan de coupe.

Tension appliqué« a chaque câble 185 t. elfe.Longueur totale des cibles 191.5kmPoids au métré de chaque cible 12.3 kgCeint, sup. + ceint, inf. 11,4 kmDalle sup. + dalle inf. 37,8 km

FIGURE 4 : Saint-Laurent 1 reactor vertical section

58

TABLE IV. HEAT EXCHANGER LEAK NUMBER

UNITSL1SL2

Vandellos

leak number391582

5. CONCLUSIONNuclear UNGG power plants today are somewhat overshadowed

by PWR reactors which, in France, produce more than 65 % of allelectricity generated at the lowest cost.

Nevertheless, there is reason to be proud of the fact thatthese UNGG reactors, with which France entered the nuclear age,have aged well despite their early flaws and discontinued commer-cial development. These flaws, steel and graphite corrosion,erosion-corrosion of heat exchangers, were compensated by anacceptable reduction of the nominal power. ISIS work showed therealism of a large-scale undertaking on upper reactor internals.

The Chernobyl event has widened the field of our safetyresearch on the UNGG reactor type. Indeed, we now need to finctthe right things to do in case of a severe accident. For instance,the use of lead would be ill-advised because it would create aneutectic with the magnesium fuel clads.

59

FORT ST. VRAIN PERFORMANCE

H.L. BREYPublic Service Company of Colorado,Denver, Colorado,United States of America

Abstract

Fort St. Vrain, on the system of Public Service Company ofColorado, is the only high temperature gas-cooled power reactorin the United States. This plant was designed by General AtomicCompany and utilizes helium as the primary coolant. The coreconsists of triso-coated uranium and thorium fuel particles castinto cylinical rods within prismatic graphite blocks. The primarycoolant system consists of the reactor, twelve steam generatormodules, and four helium circulators contained within a prestressedconcrete reactor vessel. The once-through steam generators providehigh quality 1000°F, 2400 psi main steam and 1000°F reheat steamto the main turbine. The plant has generated 3,334,000 HWH ofelectricity and has undergone 3 reactor refuel ings.

Many of the primary system components at Fort St. Vrain areproto-typical in nature and this plant has undergone an extensiveand elaborate testing program. Difficulties primarily with thehelium circulators and fuel element column movements particularlyat high power levels have resulted in significant modifications.

Power operation at 100% was achieved in late 1981 with a netthermal efficiency of nearly 39%. Excellent performance of thecoated fuel particles has contributed to a full power circulatingactivity sixty times lower than the design with subsequentclean