Studsvik Report STUDSVIK/ES-95/10 - OSTI.GOV

Studsvik Report STUDSVIK/ES-95/10

ADVANCED NUCLEAR REACTOR TYPESAND TECHNOLOGIES

Part I Reactors for Power Production

Victor Ignatiev (Editor)Olga FeinbergAlexei MorozovLennart Devell

Studsvik EcoSafe

STUDSVK ECO & SAFETY AB STUDSVK/ES-95/10

1995-02-13

Victor Ignatiev (Editor)Olga FeinbergAkxei MorozovLennart Devell

ADVANCED NUCLEAR REACTOR TYPES AND TECHNOLOGIES


Abstract

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fissionreactor types and fuel cycle technologies. Two parts have been prepared:

Pan IPann

Reactors for Power ProductionHeating and Other Reactor Applications

Pan m, which will cover advanced waste management technology, reprocessing and disposal fordifferent nuclear fission options is planned for compilation during 1995.

This catalogue was prepared according to a special format which briefly presents the project title,technical approach, development status, application of the technology, reactor type, power output,and organization which developed these designs.

Parts I and II cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and moltensalt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applicationsas power production, heat generation, ship propulsion, space power sources and transmutation of suchwaste are included.

Each project is described within a few pages with the main features of an actual design using a tablewith main technical data and figure as well as references for additional information. Each chapterstarts with an introduction which briefly describes main trends and approaches in this field.Explanations of terms and abbreviations are provided in a glossary.

ISBN 91-7010-258-9

Reviewed by Approved by

© Studsvik Eco & Safety AB, Sweden 1995

y ^

STUDSVIK ECO & SAFETY AB STUDSVK/ES-95/10

1995-02-13

Preface

For several years Studsvik AB has had d r special task, rii«gn»>pH by dr.Swedish Ministry of Industry and Commerce of surveying, analysing andreporting on die development efforts and achievements abroad, concerningadvanced nuclear fission technology. Program management and most of dietask- K |Efrfn""^ M * i " **«» hÊFy r'^fff^y SmHwrilr Era * Safety AttThe results are presented in an annual summary report as well as in topicalreports in Swedish. Contacts and collaboration widi colleagues abroad and

institutes »ml TMrareh «^»niffrfifln!t y y nf inipniianrii» in this wflfk Inrecent years, dose contacts have been established between die RussianResearch Centre "Kurchatov Institute", in Moscow and Studsvik. As a jointproject and in close co-operation k was decided to prepare a documentwhich briefly presented die plans, projects and achievements widiin die areaof advanced nuclear fission technology and dieir present status in variouscountries.

The first and second parts of this work concern nuclear fission reactors. Thelast will examine front- and back-end of advanced nuclear fuel cycles. Theintention is to extend and also update die material compiled in new editionswhen necessary. The first two parts were compiled by Victor Ignatiev(Editor), Olga Feinberg and Alexei Morozov who are leading scientists ofdie RRC "Kurchatov Institute" and I would like to tiiank diem all for theirefforts which made this report possible.

Acknowle ? * ;nts are due to Acad. N Ponomarev-Stepnoi andProf. A C t -hskifiom RRC "Kurchatov Institute" who supported thework anc / »a ML, V Krett, C Goetzman from IAEA, Stanislav Subbotinefrom RC.' S jrchatov Institute", Tor Pedersen of ABB Atom and BengtPershagt r 01 providing necessary information and useful suggestions.

I would ? r» like to thank other contributors, among diem ElisabetAppelf r for the typing and Monica Bowen-Schrire for correcting dieEngiisJ

I hope tV it the report will be of help to those who would like to obtain anoverview j f recent advancements in dus area

Lennart DevcllProgram ManagerAdvanced Nuclear Technology

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Table of Contents

Introduction

Table of Descriptio» Fi


Introduction

A Advanced Water-Cooted Reactors

AE Evolutionary designs for large power plants

AEO IntroductionAEP1 N4AEP2 SizewellBAEP3 KWU-ConvoyAEP4 APWRM/WAEP5 System 80 PlusAEP6 W E R 1 0 0 0 (V-392)AEP7 EPR

AEB1 ABWRAEB2 BWR90

AEH1 CANDU9

AT Evolutionary designs for medium power plants

ATO IntroductionATP1 W E R 500/600ATP2 AP-600ATP3 MS-600ATP4 AC-600

ATB1ATB2ATB3ATB4ATB5

ATG1

ATH1

SBWRHSBWRTOSBWR-900P1000 Natural Circulation BWRSWR-1000

MKER-800

CANDU3

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AI Innovative designs

AIO IntroductionAIP1 PIUS-600AIP2 ISERAIP3 ISISAIP4 MAPAIP5 VPBER-600AIP6 SIRAIP7 SPWRAIP8 B-500SKDIADS SCLWR

B Liquid Metal Fast Reactors

BO IntroductionBTS1 EFRBTS2 BN-600MBIS3 DEMOFBRBIS4 SAFRBIS5 AI.MR (PRISM)BIL1 BREST 300BIL2 LFBR

Gas-Cooled Reactors

COciiCI2CI3CI4

IntroductionHTR-ModulMHTGRGT-MHRHTGRMHD

D Molten Salt Reactors

DODI1DI2DI3

IntroductionUSRFUJl-PuMSR-NC


STUDSVIK ECX) & SAFETY AB STUDSVIK/ES-95/10

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Part II Heating and Other Reactor Applications

Introduction

E

F

G

District-Heating Reactors

E0ElE2E3E4E5E6E7E8E9E10Eli

IntroductionAST-500MMARSSECURE-HNHP-200Thermos-100NHR-200RUTAKNDHRSlowpoke (SES-10)GeyserGHR-20

Decentralized Nuclear Heating

FOFlF2F3F4F5F6

IntroductionCAREM-25PAES-100TRIGA Power SystemABVELENACompact HTGR Gas Turbine

Ship Reactors

GOGlG2G3G4

IntroductionOK-900AKLT-40MRXDRX



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Space Nuclear Reactors

10II1213141516TJ1819

IntroductionGPHS-RTGROMASHKATOPAZ-nSTAR-CSP-100ERATOLMCPBRRover/NERV\NPPS

Subcritical Accelerator-Driven Waste Transmutationand Energy Generation Systems

JOJlJ2J3J4J5J6

IntroductionEnergy AmplifierPHOENKJAERI-TPCATW/ABCAMSBBBR

Appendix I List of Abbreviations and Glossary of Terms


STUDSVIK ECO & SAFETY STUDSVK/ES-95/10

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Introduction

Since the beginning of the nuclear fission era, illustrated e.g. by the start-up ofthe Fermi pile in 1942, increasing efforts have been made to develop and useimproved nuclear fission reactors. The civilian application of nuclear fission ismainly power plants for electricity generation but, to a small extent, also forheat generation. There are reactors for research, radio-nuclide production,medical treatment, water desalination, ship propulsion and space applications.Also, some different systems in critical or subcritical accelerator driven modehave been studied for waste transmutation, weapon plutonium incineration,tritium production etc.

Many overviews and proceedings from recent conferences are available whichdescribe the evolution of reactor design, technologies and applications invarious countries (some of them are listed below). The present document is acomprehensive collection of short descriptions of new and advanced nuclearreactor types and technologies on a global scale (see Table 1). The purpose ofthe document is to provide an overview of present plans, projects, designs andconcepts of future reactors. The material has been collected from the openliterature and from information from the vendors.

The description of each design or concept follows a standardized format (seeTable 2) to facilitate comparison. The current status of the technology isdescribed for each design as well as type of application, reactor type, poweroutput etc.

Compilation of data began in January 1994 and the present document has beenissued as the first edition which is intended to be updated. The authors wouldtherefore appreciate any comments, corrections and additions.

References

o

o

Status of Advanced Technology and Design for Water-CooledReactors: Light Water ReactorsVienna, 1988, IAEA-TECDOC-479.

KABANOV, L, KUPITZ, J and GOETZMANN, C AAdvanced reactors: Safety and environmental considerations.An international perspective on the next generation of nuclearplants.IAEA Bulletin, 2/1992.

GAGARINSKI, A Yu, IGNATIEV, V V, NOVIKOV, V M andSUBBOTINE, S AAdvanced light-water reactor: Russian approachesIAEA Bulletin, 2/1992.

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o

PERSHAGEN, BAdvanced light Water Reactors (in Swedish)Studsvik AB, Sweden 1990 (STUDSVIK/NS-90/141).

PERSHAGEN, BNew light Water Reactors (in Swedish)Studsvik AB, Sweden 1992 (STUDSVIK/NS-92/61).

Small and medium reactors. Status and prospects. Report by anexpert group, OECD NEA, Paris, 1991.

ALEKSEEV, P N, IGNATTEV, V V, SUBBOT1NE, S A e aReactor plant designs with die enhanced safety: analysis of dieconcept (in Russian)Energoatemizdat, 1993.

PORSBERG, C W and REICH, W JWorld-wide Advanced Nuclear Power Reactors with Passive andInherent Safety; What, Why, How, and Who.Report ORNL/TM-11907, Sept, 1991.

ANGELO.JASpace nuclear power, 198S, Orbit book company, USA.

ANS Topical Meeting on Safety of Next Generation PowerReactorsSeattle, May 1988.

IAEA Workshop on the Safety of Nuclear Installations: FutureDirectionsChicago, August 1989.

IAEA Technical Committee Meeting on Technical and EconomicAspects of High ConvertersNiimberg, March 1990.

ENS/ANS-Foratom ENC90 ConferenceLyon, September 1990.

ENS TOPNUX'93 International ENS TOPical Meeting Towarosthe Next Generation of Light Water Reactors.The Hague 25 - 28 April, 1993.

Global '93 - International Conference and Technology Exhibitionon Future Nuclear Systems: Emerging Fuel Cycles and WasteDisposal Options.September 12 -17,1993,, Seattle, Washington.

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7th International Conference on Emerging Nuclear EnergySystems (ICENES '93).September 20 - 24,1993, Makuhari, Chiba, Japan.

Advanced power systems - design technology andstrategies for their deployment18-22 October 1993. Seoul Rubublk of Korea.

ARS "94 International Topical Meeting on Advanced ReactorsSafety.

17 - 21 April 1994. Pittsburg. USA-

Overview of physics aspects of Different Transmutation Concepts.1994, OECD, Report NEA/NSC/DOC(94)11.Developing Technology to Reduce Radioactive Waste May TakeDecades and Be Costly.1993, December. USA, Report GAO/RCED-94-16.

AIP Conference Proceedings 324.1995, January 8-12, USA, NM. Albuquerque.

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Tablel Advanced Reactors

P t f t l KtMlOfS wOt FtHHX

MW(e)

AE E

AEP1 N4 PWR M00

AEP2 SöeweCB PWR 12» U S A * WcNKkarBecsk

AEP3 KWU-Conoy PWR 1287

AEP4 APWR PWR 1350 Dettfleddcsgn

AEPS Syflem80PhB PWR 1345 ABB-CE ign

AEP6 WER-1000(V-392) PWR 1000 RF

AEP7 EPR PWR 1500 SioKK-KWU/ Baoc design

AEB1 ABWR BWR 13S6 O WIIM Itl l l

AEB2 BWR 90 BWR 1379 ABBAiom

AEH1 CANDU9 HWR 1050 ABCL

AT EvolatiooaryDesipH for Mcdiaa Power Plaatt

ATP1 WER-500/600 PWR 600 RF Hydroprcs Detåksl

ATP2 AP-600 PWR 630 US Detailed design

ATP3 MSÔO PWR 630 Jap» Kfimbidu Basic design

ATTH AC-600 PWR 600 Quna CNNC

ATB1 SBWR BWR 600 US Gcrtcxal Ficftnc gn

ATB2 HSBWR BWR 600 Japan Hitachi esgn

ATB3 TOSBWR-900P BWR 310 Japan Toshiba gn

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Table 1 Advanced reactors (confd)

SEATB4

A7B5

ATC1

ATH1

Al

AO>1

AV2

AIP3

ADM

A K

AIP6

AIF7

AIP8

AIP9

B

BT

BTS1

BTS2

BI

BIS3

BIS4

BUS

lOOONanmlCkaååömVWR

SWR1000

MEER-SOO

CANDU3

BWR

BWft

LM»

HWR

S K /

1000

1000

no

450

IS

Gaw

RF

C t a a *

rr—rwu

RDÖt

AECL

Banc design

Sate

Dcaäled dcspi

PIUS-600

BER

ISIS

MAP

VPBER-600

SIR

SPWR

B-5OOSKDI

SCLWR

PWR

PWR

PWR

PWR

PWR

PWR

PWR

PWR

PWR

600

200

200

300

600

320

600

515

1100

S m * .

J - P -

taiy

US

RF

US/UK

RF

Japan

ABB A n a

JABU

fVHala^HaV < f a B

Ä;OKBM

ABB-CE^tolkRojwe

JAERI

KRC-KJfHyauprcss

UancnMy of Tokyo

Saafcs

Sariia

Banc design

Basic design

Banc design

Sojdks

Stadia

Stadks

Liqaid Metal Faa Reactors

EvotetmaryDeafas

EFR

BN-600M

LMR

LMR

1500

600

Fnact/GB/GamanyAtaly

RF

EdF/NS/Bayuawut/ENEL

OKBM

Banc design

1 ailvraMBân n^^^&wk

laaorative Designs

DEMOFBR

SAFR

ALMR (PRISM)

LMR

LMR

LMR

660

450

150

Japan

US

US

FEPS

RockweO IntCE

Genenl Electric

Bapc design

Banc design

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Table 1 Advanced reactors (cont'd)

Abbre-riatioB

BIL1

BIL2

C

d l

CD

CD

O4

D

DU

DI2

DI3

Name

BREST300

LFBR

Type

LMR

LMR

She/MW(e)

300

625

Contry

RF

Japm

Vcador

RDIFE

JAERI

DeveiopBent

Studies

Studies

G u Cookd ReactofiAnoTrtive derigM

HTR-Mntal

MHTGR

GT-MHR

HTGR-MHD

USR

FUJI-Pu

MSR-NC

Ritat

HTGR

Hit»

Ritat

MSR

MSP.

MSR

80

190

300

860

P i

625

100

470

US

IS

j * r »

US

Jspin

RF

Skmen-KWU

Genenl Atomics

JAERI

ORNL

Tokii Univcnily

RRC-KI

BttK dcsgn

Studies

Studies

Studies

Studies

Studies

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Table 1 Advanced Reactors (cont'd)

Partn Heating and Other Reactor Applications

Abbre-TiatMB

E

El

E2

E3

E4

ES

E6

E7

E8

E9

E10

Ell

Name Type Sac/MW(t)

Cmmry VcatdarOnuiuliiia

Derdopment

District Heatiag Reactors

AST-500M

MARS

SECURE-H

NHP-200

Thermos-100

NHR-200

RUTA

KNDHR

Slowpoke (SES-10)

Geyser

GHR-20

PWR

PWR

PWR

PWR

PWR

PWR

LWR

LWR

LWR

PWR

HTGR

500

600

400

200

100

200

20

10

10

10-50

20

RF

Italy

Sweden

Germany

France

duna

RF

Korea

Canada

Switzerland

Germany/Switz.

OKBM

RomeUniversiry

ASEA

CEA

INET

RDIPE

KAERI

AECL

PSI

KWU

Commercial

Studies

Studies

Detailed dPsiRP

Basic design

Detailed design

Studies

Studies

Commercial

Studies

F Decentralized Nuclear Heating Power Reactors

Fl

F2

F3

F4

F5

F6

G

Gl

G2

CAREM-25

PAES-100

TRIGA Power System

ABV

ELENA

Compact HTGR GasTurbine

PWR

PWR

PWR

PWR

PWR

HTGR

100

160-170

64

60

3

29

Argentina

RF

US

RF

RF

US

INVAP

OKBM

General Atomic

OKBM

RRr.KI

General Atomic

Studies

Commercial

Commercial

Detailed design

Bask design

Studies

Ship Reactors

OK-900A

KLT-40

PWR

PWR

170

170

RF

RF

OKBM

OKBM

Commercial

Commercial

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Table 1 Advanced Reactors (cont'd)

Abbrt-

G3

G4

I

11

Q

D

M

15

16

n

18

19

J

Jl

J2

J3

J4

J5

J6

Naaae

MRX

DRX

GPHS-RTG

ROMASHKA

TOPAZ-n

STAR-C

SP-100

ERATO

LMCPBR

Rov«/NERVA

NPPS

Type

PWR

PWR

doctnc

Fan.

ckrnn

LMR,

Fatt.ThfTTnknuc

LMR7HTR

LMR/HIR

HTGR

H I T »

Sat/MW(t)

100

0.75

Inwiriai

3-HHMWe

MWe

6-10-JMWc

MWe

0.1 MWe

210"2MWe

10

367-1566

1200

Sabcritieal Accelerator Drives Waste Traawtatioa i

Energy Amplifier

PHOENIX

JAERI-TPC

ATW/ABC

AMSB

BBR

LWR/HTGR

LMR

MSR/LMR

MSR/LMR

MSR

MSR

200

3600

820

3000

1000-2100

5000

Coaatry

Japan

J a p -

US

RF

RF

US

US

Fiance

Japan

US

RF

MdEaerty Genen

Swioeriand

US

Japan

US

Japan

Pnnct

Vendor

CXfAuatua

JAERI

JAERI

w . ^ ^

RRC-K1

RRC-K1

Rockwell Insände

(fffnrral FJenric

CNES/CEA

JAERL Toshiba

LANL,

, ifgnmiff

RRC-H

CERN

BNL

JAERI

LANL

Tokay University

CEA

Ptyclopmcnt

Bane desgn

Detaileddesign

Basic deä

Batiedesini

Stadia

Stadia

Ran* ilfjuun

Stadia

Stadia

Stadia

Stadia

Stadia

Stadia

Stadia

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Table 2 Advanced Nuclear Design Description Format

Format Description

Technical approach

Title

Applicaton of the reactor

Reactor type

Power output

Organization

Development status

Description

Schedule

References

E Evolutionary design for large power plantsT Evolutionary design for medium power plantsI Innovative design

Title of project

NPP Nuclear power plantNHP Nuclear heating plantNHPP Nuclear heat and power plantSMS Ship mobile systemSNS Space nuclear sourceTWS Transmutation waste system

PWR, BWR, HWR, LWGR, FBR, HTGR, MSR, SADS(subcritical accelerator-driven system) e.a.

L Large power (more than 1 000 MWe)M Medium power (more than 100 MWe, less than 1 000 MWe)S Small power (more than 1 MWe, less than 100 MWe)VS Very small power (less than 1 MWe)

Type of organization, which developed design

C Commercially available design

D Detailed engineering design

B Basic engineering design

S Studies of concept

General description of the design. This will generally include asimple description of the reactor plant. This may be followed by asummary with the main features of an actual design using tableswith main technical data or figures

Schedule for development

References for additional information.

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Part I Reactors for power production

Introduction

In many countries, nuclear power has demonstrated its capacity to generateelectricity on a large scale, cost-efficiently and with low environmental impact

Of the current 430 operating NPP units, almost 350 are LWR ones. Theoperating experience of these plants shows that they represent an economic andreliable source of electricity production. They also represent a broadlydeveloped and mature basis of technology for further nuclear power develop-ment The objectives for improvements cover a broad range of interest such asrisk minimization, lower energy cost production, improved reliability andenhanced safety.

Progress and schedule for the development and deployment of ALWR

After the severe accident at Three Mile Island Unit 2 authorities in severalcountries requested measures for enhanced safety at operating plants. Amongthese measures, filtered venting of the containment was installed in many plantsin order to mitigate consequences in the event of some specific severe accidentscenarios.

The development of the ALWR Program started in the USA in the beginning ofthe 80's on the initiative of EPRI supported by utilities, DOE, US reactorvendors and architect engineering firms.

This program provided a foundation for the Nuclear Power OversightCommittee (NPOC) comprehensive initiative for revitalizing nuclear power inthe USA, as set forth in its "Strategic Plan for Building New Nuclear PowerPlants", published in November 1990 and updated annually each November.

Work on the Utility Requirement Document, URD, started in 1985 and wascompleted in 1990. The cooperative efforts, First-of-a-Kind Engineering,(FOAKE) between the US utilities, DOE and the vendors started in 1992.

In the USA the licensing process was changed in 1994. An important newfeature of this process is the issuing of a final design approval (FDA).

The time schedule for safety evaluation, FDA and FOAKE is given in thefollowing table, reproduced from an IAEA document In fact, the FDA forABWR and System 80 Plus was issued in mid 1994 not far behind schedule.

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ABWR System 80+ AP-600 SBWR

Final Safety Evaluation 5/94Report

Final Design Approval 6/94

6/94

8/94

First-of-a-KindEngineering

9/96

-4/96

-6/96

9/96

-3/96

-5/96

After the severe accident at Chernobyl Unit 4, the IAEA requested itsInternational Nuclear Safety Advisory Group INSAG to give recommendationsfor action to enhance reactor safety. In 1988 the IAEA issued the INSAG-3report with the title of Basic Safety Principles for Nuclear Power Plants. Thiswas followed by further reports:

o INSAG-4o INSAG-5o IAEA Safety Fundamentals

Safety Culture (1991)The Safety of Nuclear Power (1992)The Safety of Nuclear Installations (1993)

Already in the INSAG-3 report from 1988, a distinction was made betweenoperating plants and future plants. For future plants, the probability of severecore damage should be less than 1 for every 100 000 reactor years and theprobability of large releases to the environment should be less than 1 for every1 million reactor years.

A similar development effort for a European Pressurized Water Reactor (EPR)was launched in Europe between France and Germany in early 1992 followed bythe start in 1992 of producing a European Utility Requirement Document,(EUR). The first two volumes of the EUR were published in February 1994 andthe main part of the remaining document is scheduled for completion by the endof 1994.

The advanced nuclear power plant designs currently being developed representa wide range of alternatives. Most of the reactor concepts are evolutionary andrepresent small or moderate extensions of current designs. Some of them useupgrades of operating plants. Few concepts based on well-known LWR, HWR,LMFBR technologies have significant departures from the current ones. Thereare also some new designs that use innovative principles, which are uniquecompared to current practice, e.g. the PIUS principle for LWR, modular hightemperature gas-cooled reactors (HTR), liquid metal reactors (LMR) andmolten salt reactors (MSR).

The typical designs belonging to water-cooled, LMR, HTR and MSR groupsare presented in Table 1 (see p 4).

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12

In this catalogue, reactor design improvement follows three directions (seeFigure 1, p 14) described below.

Evolutionary designs for large power plants

A design approach involving "step by step" modifications of present-daydesigns, with an emphasis on design proveness. Here, more traditional designersfind smaller ways to modernize design to meet more stringent new standards.These (mainly large power - more than 1 000 MWe) provide the improvementof NPP reliability and operating safety with the simultaneous improvement ofeconomic features due to further improvement and simplification, to someextent of the design. These designs, which incorporated extra safety features,require only small additional testing.

Evolutionary designs for medium power plants

A design approach which allows for the improvement of the NPP reliability andsafety with simultaneous improvement of the economy and competitiveness dueto simplification of the NPP design and reduction of specific core power and theunit power of a reactoi (mainly about 600 MWe), the construction period, thedegree of financial risk and expected broadening of the market in the developingcountries. This approach refers to the evolutionary plant with significantinnovative features, which are, however, within the range of existing technicalknowledge.

Innovative designs

Some of the new ALWR designs ensure the improvement of LWR-based NPPand NHP reliability and safe operation mainly due to changes in the designarrangement and performances or in plant responses to the accidents. The majordevelopment goals stated for some of these plant designs are simplification andan enhanced use of passive safety features, based on current plant designtechnology. Others, mainly non-ALWR reactor types, have more ambitiousdevelopment goals that, in turn, require more radical modifications of designaiming at the maximum possible utilization of inherent safety features and self-protection capacity.

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References

TAYLOR, J J and SANTUCCI, JSafety, technical, and economic objectives of EPRFs advancedlight water reactor program.Symposium on Advanced Nuclear Power Systems.Seoul, Republic of Korea, 18-22 Oct, 1993 (IAEA-SM-332/II.l)

Review of advanced LWR design approaches, 1994,1AEA-TC-879.


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13

BACHER, P and BOARD, JIAEA-SM-332.

GOLAY.MWAdvanced Fission Power Reactors.Ännu Rev Part Sci, 1993, v 43, p 297-332.

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c

so.B

COco

I

ABWRSizeweli B[ lBWR 9Öl [ÄPWR W/M

Convoi EPR WER-92N4l [System 80+1 [Candu 9

CurrentLWR

\ Large sizeALWR

MS-600

WER-500 AP-600

Candu 3| lAC-600

MKER-8001 1 SBWR 1 iTSBWRl

HSBWR

Mid sizeALWR VPBER MHTGR

ALMR

SIR MARS

PIUS-PWR, BWRModular HTR

Modular LMR (Na, Pb)MSR, other

Figure IDifferent approaches to improvements of advanced fission reactor designs.

Innovations

C/3

m3

i - - C/3

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A Advanced Water-Cooled Reactors


AEO Introduction

The designs of new light water reactors have to meet criteria concerningimproved safety, economic competitiveness and user friendliness. Criteriahave been issued by utility groups and authorities, which have beenmentioned earlier.

The evolutionary approach has the benefit of providing the following:

oooooo

Predictable performance based on experienceEnhanced safety, performance and economicsIncremental changes in technologyAvailability of proven equipment and systemsAvoiding need to additonal experimental verificationAssured acceptance and licensibility

In comparison with the first stage LWR generation evolutionary ALWRdesigns underwent a considerable change. The change involved, basically, theenhancement of a power of units, thermodynamic parameters of a steam-turbine cycle, fuel utilization, operational reliability and safety.

The safety of evolutionary power plant designs has been improved on arealistic and efficient basis without jeopardizing economics.

In summary, the following is a list of the most significant safety improve-ments applicable to one or more reactor types:

o Reduction of the core power density

o Increase in the number of control drives

o Application of spectral controls to compensate the variation ofreactivity with burnup

o Increase in the cooling water inventory capable of being storedwithin the containment and primary circuit space

o Application of the emergency core cooling system (ECCS) toeach loop

o Improvement of NPP protection in case of a loss of off-sitepower

ww victor/ A-awcr ea


1995-02-13

A:2(3)

o Improvement of off- and in-rcactor monitoring for predictingthe reactor power density and temperature distributions on areal time basis

o Reduction of the neutron fluence on the reactor vessel

o Mitigation measures for severe core damage accidents with

failure of the reactor pressure vessel (RPV)

o Filtered containment venting

o Application of a double containment protected from externaleffects by strong reinforced concrete structure

o Replacement of block-type control boards with multifunctionaldisplays

o Division of the personnel's function among the operatorsresponsible for maintaining rated operating conditions,engineers responsible for NPP control under emergencyconditions, operators responsible for maintaining reactoravailability during outage, refuelling, repair

o Increased use of periodical simulator training of operator,maintenance personnel etc

o Development of systems capable of checking the mainequipment operability during reactor operation

This design approach obviously implies a step-wise evolution of the plantdesign on the basis of the utility requirements and stated preferences. Largepower ALWR designs are now under consideration in France, Germany, UK,USA, Japan, Russia, Sweden and some other countries. Canadian HWR linesare also presented.

References

o

GAGARINSKLAYuetalAdvanced Light Water Reactors: new ideas and approaches.Nuclear Society International, Moscow, RF, 1992.

The new reactors, Special session.Nuclear News, 1992, September, p 696-90.

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1995-02-13

A:3(3)

o PEDERSEN.TJCurrent trends ii? safety philosophy and design goals foradvanced nuclear power plant designs.Proceedings ASME/ISME Nuclear Engineering Conference,Vol 2,1993, ASME, p 539-544.

Survey of containment designs for new/advanced waterreactors.1994, OECD, Report NEA/CSN1/R(94)5.

ww victor/A-awcr ea


1995-02-13

AEPl:i(3)

AEP1

Format

Title

Application of thereactor

Reactor type

Power output

Organization (name)

Development status

Description

Description

N4

NPP

PWR

Large powei

Electricité d

Commercial

The French

Schedule

The French N4 plant is a 1400 MWe PWR with 4 train systems organizabon.The N4 is fuelled by enriched UO2 pellrts and uses Zircalloy-4 as a claddingmaterial. The active core height is 4.2 m and it contains 205 fuel assemblieswith a 17x17 array. As regards safety, it was designed in an evolutionaryperspective, benefiting the experience gained from the previous standardizedseries of 900 MWe and 1300 MWe plants, and using technological progressand new developments after careful validation.

The safety improvements are the following:

0 More accurate and functionally adapted instrumentation

0 A better global methodological approach for fire hazards

0 Means to deal with total loss of frequently used systems, andwith shut-down conditions

0 An increased role for PSA in the design and licensing process

Other important features of N4 plant arc a computerized control room andmitigation of core melt in ultimate conditions, with filtered ventingcontainment as last option.

The N4 three units are under construction in France. The first N4 unitChooz B-l is to be commissioned in 1995.

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STUDSVIK ECO & SAFETY AB STUDSVK/ES-95/lö

1995-02-13

AEP1:2(3)

N4 Unit Data

Thermal power

Reactor CoreEquivalent diameterNumber of control assembliesInitial enrichmentEnrichment at equilibriumAverage fuel burn-up at equilibriumTotal weight of UO2

Reactor Coolant SystemOperating pressureReactor vessel inlet temperatureReactor vessel outlet temperatureNumber of reactor coolant pumps

Reactor Pressure VesselInside diameterTotal heightDesign pressureDesign temperature

ContainmentConfigurationGross volumeDesign pressure

Steam GeneratorsNumberSteam pressure at SG outletSteam temperature at SG outlet

4 270MW

3.47 m731.8/2.4/3.1 % U-2353.4 % U-23539MWd/kg110.41

155 bars.292.2 °C329.6 °C4

4.5 m13.64 m172 bar343 °C

Double72000 m3

5.3 bar

472.3 bar288 °C

References BERGER-PERRIA, M M (EdF) and LANGE, M (Framatome)"Basic Information on Design Features of the N4 Nuclear Power Plant"Review of Advanced Light Water Reactor Design Approaches, Moscow,RF, 10-13 May, 1994, IAEA-TC-879.

ww victor/AEP1 ea

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n V l . l ' l

tt.di


1995-02-13

AEP2:i(3)

AEP2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

Sizewell B

NPP

PWR

Large Powc

Westinghou

Commercial

The reactor i

Schedule

ww victor/AEP2 ea

Westinghouse/Nuclear Electric, USAJnited Kingdom

The reactor design is developed on the basis of well-established PWRpractices, derived from the Westinghouse SNUPPS (Standardised NuclearUnit Power Plant System) design.

The fuel is low-enriched uranium oxide claded in Zircaloy 4. The corecontains 193 fuel assemblies, consisting of 236 fuel rods each. Control rodsforming rod cluster control assemblies can be inserted into the core ininterstitial spaces within certain of the fuel assemblies and they are groupedinto a number of "banks" for reactor control and for shutdown.

The core is cooled by water at a pressure of 15.5 MPa. The reactor coolantsystem comprises the reactor pressure vessel, four centrifugal reactor coolantpumps, four vertical steam generators with the primary cooling circuitpassing through an inserted U-tube bundle, a pressurizer and linkingpipework.

The important engineering safety features of the reactor plant are:

o 4-train system organization

o 30 minutes tolerance before operator action

o Fully computerized reactor protection system

o Station black-out addressed by two reliable g r i s and fouremergency systems

The reactor design incorporates a considerable degree of redundancy anddiversity in systems providing safety functions.

The plant is owned and operated by Nuclear Electric, the electrical utilityoperator of nuclear power stations in England and Wales, with scheduled


1995-02-13

AEP2:2(3)

start-up in 1994. Nuclear Electric has formally applied for planning consentto build Sizewell C unit which would replicate the Sezewell B design.

References

Sizewell B Unit Data

Thermal capacity

FuelMaterialEnriched (feed fuel)Fuel clad material

Reactor coreActive heightEquivalent diameterMass of UO2 in the core

Reactor vesselOverall heightInside diameterMaterialInternal claddingDry weight

Steam generatorNumber

Turbo generatorNumber

ContainmentConfigurationGross volumeDesigned pressure

3 411 MWt

sintered UO23 . 1 %Zircaloy4

3.66 m3.37 m1011

13.52 m4.394 mlow alloy steelstainless steel435 t

4

2

Double91000 m3

3.5 bar

BOARD, L A and QUICK, M VThe Sizewell B PWR Design,Review of Advanced Light Water Reactor Design Approaches, 10-13 May,1994, RF, Moscow, IAEA-TC-879.

Nuclear Electric Ann<~ jnces Sizewell C.Atom, 1993,431, November/December, p 10.

ww victor/AEP2 ea

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f f i ;'." iJ --

' 1i

i

'i

i

!

j

f

Control rod drivemechanism

Thermal.sleeve

Upper support.plate

Internals supportledge

Core barrel

Closure headassembly

Lifting lug

Outlet nozzleUpper core

plate

Reactor vessel

Core supportcolumns

Inlet nozzle

Fuel assemblies

Irradiationspecimen guide

Lower coreplate

Neutron shield

Bottom supportforging

Radial support

Lowerinstrumentationguide tube

I i m i i v \ l l ' 2S i / - A •!! !', K M , i . . . I ' ; • , i i • \

STUDSVnC ECO & SAFETY AB STUDSVK/ES-95/10

1995-02-13

AEP3:l(3)

AEP3

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

KWU - Convoy

NPP

PWR

Large power/1287 MWe

Siemens - KWU, Germany

Commercial

The KWU-Convoy Nuclear Steam Supply System (NSSS) consists of a PWRwith four primary coolant loops, a pressurizer connected to one of the loops, 4steam generators, four reactor coolant pumps and the auxiliary and safetysystems directly related to the NSSS. The NSSS generates approximately3 782 MWt producing steam at 63.5 bar at the steam generator outlet. Theturbine generator provides a net power of approximately 1 287 MWe. Fullload rejection is accepted without a reactor or tubine trip. The turbine plant iscompletely automatic and supervised from the control room.

A short summary of main safety and operational features is given below.

The reactor core concept includes:

o An RPV with a wide gap between the core barrel and RPV

shell which keeps the integral neutron dose to the RPV beltlineat a low level Qimitation of embrittlement).

o Low linear heat generation rate, resulting in high bumupcapability and increased fuel management flexibility.

o The use of Zircaloy 4 guide thimbles and spacers increases

neutron economy.

o Gadolinium poisoning of fuel rods improves fuel utilization.

The reactor building concept includes:

o The use of the spherical steel containment, designed for fullpressure and temperature after a double-ended LOCA.

A fuel pool inside containment, which facilitates handling andshortens refueling time.

ww victor/AEP3 ea

STUDSVUC ECO & SAFETY AB STUDSVK/ES-95/10

1995-02-13

AEP3:2(3)

Schediile

o The use of a reinforced concrete containment protects theNPP against external events.

The last series of PWRs built in Germany is the standardized type of KWU-Convoy concept However, there is always room for further improvements inthe technical field.

KWU-Convoy Unit Data

Thermal power

Fuel assemblyArrayNumber of fuel rodsNumber of guide tubes for absorber

Fuel RodLengthOutside diameterCladding material

Fuel pelletEnrichment for the first coreFor the reload coreAverage fuel bumupTotal weight of the UO2

Reactor coolant system

Design conditionsPressureTemperature

Reactor coolant pumpTypeNumber

Steam generatorTypeNumber

ContainmentConfigurationDiameterDesigned pressure

3 782 MWt

18 x 18-2430024

4.4 m9.5 mmZircaloy4

1.9; 2.5; 3.2 %3.2; 3.4 %31.8MWd/kg1031

175 bar350 °C

Single-stage ce4

U-tube, vertica4

Double56 m5.3 bar

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1995-02-13

AEP3:3(3)

References CZECH, L and FE7GEL, A etc.Technical Information on Design Features of Siemens Convoy PWRReview of Advanced Light Watei React jr Design Approaches.10-13 May, 1994, RF, Moscow, IAEA-TC-879.

Section

1 Reactor pressure vessel2 Refuelling machine3 Lay down position for

core internals4 Fuel pool

5 Reactor building crane 9 Gantry6 Pressurtzer 10 Main steam and feedwater valve room7 New fuel store 11 Pipe duct8 Equipment lock 12 Cable duct

Figure AEP3KWU-Convoy: containment

ww victor/AEP3 ea


1995-02-13

AEP4:i(4)

AEP4

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

APWR (Advanced Pressurized Water Reactor)

NPP

PWR


Mitsubishi/Kansai Electric Power Co (KEPCO

Detailed design

The APWR is based on the design, which incoisimplification and additional operating margins, and operating experience ofcurrent PWRs. Basically, the APWR is a four-loop PWR. The thermal powerof the reactor is 3823 M W L The fuel rods are assembled in square arrays.Compared with Sizewell B and KWU-Convoy plants (see AEP2 and AEP3),the number of fuel assemblies is unchanged, but the core density is loweredby 20 % relative to conventional designs, by using larger size assemblies.Each fuel assembly consists of a 19 x 19 array of pins cladded by Zircaloy 4(instead of a 17 x 17 traditional array). A fuel assembly contains 16 thimblesto house spectral control by introducing water-displacing rods. Thedisplacers remain inserted into the fuel assembly during the first half of thefuel element lifetime. At the core level in the fuel assembly, a large number ofZircaloy 4 grids is used. To enhance the mechanical strength of the fuelassembly, the fuel element claddings and grids are thickened. To reduceneutron leakage, the core is enclosed by a steel neutron reflector. With 3 %fuel enrichment in U-235, the period of continuous power operation is 13.5months. The 10 % cost reduction of the APWR fuel cycle results from abovenoted features. The major part of these improvements is due to a spectralcontrol. With "grey" control rods replaced by conventional ones, the reactorcan operate with a mixed UO2-PUO2 fuel.

Other typical features are:

0 The increased size of the RPV(all large nozzles are now wellabove the core top)

0 The containment with an emergency water storage tank at thebottom and the four-train arrangement of emergency coolingpumps (with "pairs" of diversified design)

ww victor/AEP4 ea

STUDSVnC ECO & SAFETY AB AEP4:2(4)

1995-02-13

Schedule

The steam generator and turbine designs are improved. To reduce theradiation dose of the personnel, access is improved to the units which requireto repair work. Low cobalt containing structural materials are then applied.The size of the hatches for steam generator examination and repair isincreased.

The designation APWR is not generally known in the US. Westinghouse hasnot used the US certification process. Some design activities have continuedin Japan, but a decision on construction or siting has not been taken yet.Also, because utilities expressed a desire for an APWR at a nominal1 000 MWe rating, Westinghoase developed the nominal 1 050 MWe threeloop APWR.

APWR Unit Data

Electrical output

Reactor coreNumber of fuel assembliesActive heightEquivalent diameterTotal weight of UO2Number of rod cluster controlassembliesAverage power density

Reactor coolant systemPressure at vessel outletFlow rate

Reactor vesselInside diameterOverall height

Reactor coolant pumpSpeedWeight

Steam generatorHeat transfer surfaceOutlet steam pressure

ContainmentConfigurationDiameter

1350 MWe

1933.9 m3.98 m119.21

6980kW/l

157 bar88 000 m3/h

5m16 m

1 185 rpm92 t

6 039 m2

69 bar

Double60m

ww victor/AEP4 ea

STUDSVK ECO & SAFETY AB STUDSVnC/ES-95/10

1995-02-13

AEP4:3(4)

References Technical Information on Design Features of the APWR.Review of Advanced Light Water Reactor Design Approaches.10-13 May 1994, Moscow, RF, IAEA-TC-879.

The Westinghouse APWRNuclear News, 1992, September, p 74-75.

Spray Header

CoreRefloodTank

Spray Header

Spray Header

CoreRefloodTank

Spray Header

Emergency WaterStorage Tank

_JH : High Head SI Pump (4 sets)R : RHR/CV Spray Pump (4 sets)

Figure AEP4APWR: Integrated safeguards system layout around containment vessel.

ww victor/AEP4 ea

STUDSV1K ECO & SAFETY AB STl.'DSVIK/FS-y.VK)

995-02-13

AEP4:4i4)

Control Rod Drive

Mechanism

Guide Tube

Fuel RodRadial Reflector

Figure AKI'4APWR: Reactor pressure vessel internals.


1995-02-13

AEP5:i(4)

AEP5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

System 80 Plus

NPP

PWR


ABB-CE Nuclear Power, US

Detailed design

System 80 Plus is an advanceenhancements of the System 80 standardized design. The System 80 plantdesign is used for Palo Verde units in Arizona, the USA. The enhancements,as integrated design process, address new licensing issues, safetyimprovements, improved operability, availability and maintainability, plantsimplification and cost reductions.

The thermal power of the reactor is 3817 MWt. The reactor vessel isdesigned to contain and support the core and the fuel. The design of ihe coreis based on that one of System 80. The reactor vessel is a vertically mountedcylindrical vessel with a hemispherical bottom attached to the vessel and aremovable hemispherical upper closure head. The reactor vessel is fabricatedfrom low alloy steel and the internal surfaces that are in contact with thereactor coolant are cladded with austenitic stainless steel. The reactor coreconsists of 241 fuel assemblies and 94 or more control elements. Anadvanced burnable absorber strategy is implemented to simplify reactivitycontrol and provide for extended fuel burnup. Full-strength control elementassemblies consist of a Inconel clad with boron carbide or silver-indium-cadmium absorber rods. Reduced strength control rods composed of solidInconel provide the capability to change operating power level using controlrods only. This simplifies reactivity control during plant load changes andreduces liquid waste processing requirements that normally accompanychanges in soluble boron concentration.

Like previous ABB-CE reactors, the Reactor Coolant System (RCS) forSystem 80 Plus has two loop configurations. Each loop consists of one hotleg, one steam generator, two cold legs and two pumps.

A press urizer with an increased volume (in specific volume, the increaseamounts to 33 %) to reduce the pressure changes during transients such as

ww victor/AEP5 ea

STUDSVIK ECO & SAFETY AB STUDSVlK/ES-95/10

1995-02-13

AEP5:2(4)

Schedule

reactor trip and load rejection. The increased secondary side water inventory(by 25 %) and heat transfer surface of the steam generator are also aimed toimproved operating performance.

The most significant changes are in the engineering safety systems, inparticular, in the Safety Depressurization System (SDS) and in the SafetyInjection System (SIS). The SIS has been made simplier, more reliable andthe performance has been improved. It incorporates four-train safetyinjection, an in-containment refuelling water storage tank, and direct vesselinjection. The containment structure is a steel sphere enclosed in a cylindricalconcrete building. Gas turbines as a diverse AC source are used.

A PRA indicated a reduction in the risk of severe accidents by two orders ofmagnitude compared to that with previous systems.

No System 80 Plus plant has been ordered yet The final design approval forSystem 80 Plus was issued in 1994. Also, construction of System 80 NPPswith some System 80 Plus design features is now underway in the Republicof Korea.

System 80 Plus Unit Design

Electrical output

Fuel assemblyArrayNumber of fuel rodsClad materialEnrichment levelsFuel outside diameter

Reactor coreNumber of fuel assembliesCore height (active fuel)Core diameter

Reactor coolant systemDesign pressureDesign temperatureReactor inlet temperatureReactor outlet temperatureNumber of coolant pumpsNumber of steam generators

1345MWe

16 x236

16

Zircalov43.2, 2.8, 1.9 %9.7 mm

2413.813.65

17034329132342

mm

bar»C°COC

ww victor/AEP5 ea

STUDS VIK ECO & SAFETY AB STUDSVK/ES-95/10 AEP5:3(4)

ContainmentConfigurationCross volumeDesigned pressure

Double95 930 m3

3.7 bar

References Technical information on design features of System 80 Plus.Review of Advanced light Water Reactor Design Approaches.10-13 May, 1994, RF, Moscow, IAEA-TC-879.

TURK, R S and MATIZTESystem 80 +™: PWR technology takes a major step up the evolutionaryladder.Nuclear Engineering International, 1992, November, p 15 - 24.

The ABB Combustion Engineering System 80 +Nuclear News, 1992, September, p 68-69.

ww victor/AEP5 ea

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1995-02-13

AEP6:i(3)

AEP6

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

WER-1000(V-392)

NPP

PWR


EDO Hydropress, RF

Detailed design

A number of advanced \with an electric output 1 000 - 1 100 MWe are now under development inEDO Hydropress.

The following measures aimed at safety improvement in comparison withthe commercial reactor plant V-320 have been realized:

o

o

o

o

o

Advanced steam generators with a large water inventory

Four safety train design principle

The reactor coolant pump with the design of sealingsexcluding their leaks in the event of long-term blackout andthe absence of sealing water supply

Realization of a diagnostic system set and leak-bcfore-breakapproach

A filtered double containment with reinforcement of the floor

Containment filtered venting system

System of returning the primary circuit coolant

The design lifetime of the plants has been increased up to40 years.

The V-392 and V-413 as well as the V-320 commercial reactor plantshave 4 horizontal steam generators. The V-410 version introduces 4vertical steam generators.

ww victoryAEP6 ea

STUDSVIK ECO & SAFETY AB STUDSVnC/ES-95/10

1995-02-13

AEP6:2(3)

Schedule

All advanced WER-1000 reactors use the same fuel rods (diameter9.1 mm and height 3.54 m) and practically the same fuel assembly design.

Compared with V-392 and V-413, the latest V-410 version has a decreasedvolumetric core power density (from about 110 to about 85 KW/1), increasedRPV diameter and as a result reduced RPV neutron fluence.

The V-413 version, developed with FVO Engineering International (Finland)is based only on proven V-320 safety system technology. To improve thesafety the V-392 and V-410 versions incorporate additional wateraccumulators for emergency core cooling and passive residual heat removalsystems via steam generator with air coolers outside the containment

No advanced WER-1000 plant versions have been ordered yet The V-413version has been one of contenders for NPPs in Finland. Especially for theV-410 version significant verification testing and detailed design workremains to be done, prior to commercial utilization.

VVER-1000 Unit Data (V-392, V-413)

Primary pressure

Primary temperature

Secondary pressure

Steam temperature

Nominal coolant flow rate

Average linear heat rating at fuel elements

Humidity at the outlet at full power

Number of control members

Number of assemblies

Average fuel burnup fraction, (at stationaryfuel cycle)

Time of operation at rated power fora year (effective)

157 bar

320 °C

63 bar

278 °C

88 000 m3ft

166.7 W/cm

0.2%

97

163

44MWd/kg

7 000h

References FEDOROV, V G et alDesigns of advanced VVER-1000 reactors.Proceedings International Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 25-29,1992, Tokyo, Japan, v 4, p 4.4.3.

ww victor/AEP6 ea


1995-02-13

AEP6:3(3)

Pressurizer

/ e ^

Figure AEP6V-410: Reactor coolant system.

ww victor/AEP6 ea

STUDSVIK ECO & SAFETY AB STUDSVnCES-95/10

1995-02-13

AEP7:i(3)

AEP7

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

EPR (European Pressurized Water Reactor)

NPP

PWR


Siemens - KWU, Germany; Framatome, Fra

Basic design

The EPR represents an advanced, evolutions

Schedule

has been developed by Nuclear Power International (a company formed byFramatome and Siemens-KWU), based on N4 plant and KWU-Convoy plantdesigns. The design goals for EPR include improving functional safety bysystem design simplifications, strengthening of redundancy by strict physicalseparation, extending the grace period (up to 24 h) by additional and enlargedwater storage capacity. Special attention is given to severe accidents involvingcore melt corium spreading; strong full pressure double containment formargins and no need for early heat removal from containment

The thermal power of the reactor is 4 250 M W L The core is based on N4fuel with 17 x 17-25 rods per assembly. The core control principles arebased on a combination of the Siemens-KWU and Framatome experience.

The containment consists of a prestressed concrete cylinder which can befitted with a steel liner. To provide a dual wall containment function, thecontainment is surrounded by a second non-prestressed concrete cylinder,forming an annular secondary containment Safety systems are installed in anannular building surrounding the outer containment. Both the reactorbuilding and the annular building will be built on the same foundation mat(for seismic protection reasons). Separation between the four safety divisionsand one operational division is accomplished in a sector arrangement Thelower area of the containment houses the water supply for the safetyinjection systems, the in-containment refuelling water storage tank.

The development of the EPR has not yet been completed. French andGerman authorities have agreed upon the harmonization of licensingrequirements. According to the overall time schedule, the Basic Design shallbe completed during 1995 with an application for construction licensescheduled for the end of 1995. The target construction data for FOAKE is1998.

ww victor/AEP7 ca


1995-02-13

AEP7:2(3)

References

EPR Unit Data

Electrical output

Reactor coreNumber of fuel assembliesActive lengthTotal fuel assembly lengthNumber of control rodsVessel inlet/outlet temperatureEnrichment (max)Average fuel bumup

Primary and secondary systemsRCS operating pressureRCS design pressureMain steam pressure at hot standbySecondary side design pressure

Reactor pressure vesselFluencr (design-target-60 years)

1500MWe

264420 cm480 cm81291 °C/3264.9 % U23560MWd/kg

155 bar176 bar84 bar91-94 bar

1101 9n/cm2

Steam generatorHeat transfer surface (with economizer) 7 300 m2

Tubemateiial Incalloy800Water amount of 2-nd side 75 t

ContainmentConfigurationGross volumeDesigned pressure

Double75 000 m3

7.5 bar

WON, M and KRUGMANN, U et alBasic Information on the Design Features of the EPR.Review of Advanced Light Water Reactor Design Approaches10-13 May, 1994, Moscow, RF, IAEA-TC-879.

WATTENAU, M and SEIDELBERGER, HThe European PWR - a progress reportNuclear Engineering International, 1994, Oct, p 70-74.

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1995-02-13

AEP7:3(3)

Figure AEP7EPR: Containment building.

ww victor/AEP7 ea


1995-02-13

AEBl:i(3)

AEB1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

ABWR

NPP

BWR


General Electric, US; Hitachi/Toshiba Tokyo Electric Power Co (Tepco),Japan

Commercial

The ABWR with the thermal power output of 3 926 MWt is a combinedapplication of the technical solutions already proven individually in differentBWRs. The main features of the ABWR are:

o Advanced fuel and core design

o Internal recirculation pumps, at the bottom of the core

o Decreased hydraulic resistance of RCS. Lower recirculationflow pumping power

o Fine motion control rod drives with diversified scram actuation

(hydraulic piston drive + electric motor drive backup)

o Digital and solid state control

o Multiplexing

o Improved reinforced concrete primary reactor containment andreactor building

o Three emergency core cooling systems (ECCS) divisions.

The core and fuel designs aim at improving operating efficiency, operabilityand fuel economy by means of PCI (Pellet Clad Interaction)-resistant fuel,axially- zoned enrichment, control cell core design and increased core flowcapability. The latter allows for hydraulic spectral shift operation to provideadditional burnup towards the end of the operating cycle. The reactorpressure vessel has been designed to reduce the number of welds and to

ww victor/AEB 1 ea

STUDSYIK ECO & SAFETY AB STUDSVnC/ES-95/10

1995-02-13

AEB1:2(3)

Schedule

References

permit maximum in-service inspections of welds by automatic equipment Alllarge pipe nozzles to the vessel below the top of the active core have beeneliminated, improving safety performance and reducing ECCS capacity.

Two ABWR units are under construction in Japan. These units are scheduledto be taken into operation in 1996 and 1997. The ABWR is also undergoinga design certification process in the US. It has been selected by US utilities asone of two designs in a FOAKE program of 1996 that is supported by theutilities and the US DOE.

ABWR Unit Data

Electrical output

Reactor vessel pressure

Core power density

Number of fuel assemblies

Number of control rods

Core height

Core diameter

FuelAverage UO2 enrichmentNumber of fuel rodsFuel rod arrayCladding materialFuel burnup

Reactor coolant systemInlet feedwater temperatureOutlet feedwater temperatureNumber of pumps

TurbineNumberPeak electric powerInlet steam pressureInlet steam temperature

1 356 MWe

71.6 bar

50.6 MW/m3

872

205

3.71m

5.16 m

3.2%628 x 8Zircaloy232MWd/kg

215.5 °C287.4 oC10

11381 MWe67.5 bar283.7 °C

Safety Aspects of Designs for Future Light Water Reactors(Evolutionary Reactors)IAEA, Vienna, 1993, IAEA-TECDOC-7123.

GE's ABWR and SBWRNuclrar News, September, 1992, p 70-74.

ww victor/AEB 1 ea

STUDSVIK ECO& SAFETY AB STI'DSVIK HS-95/11

1995-02-13

AEB1:3(3)

1J

:\

1

;")

fi

t

8

<)

10

1 1

IL'

i:?>

1 1

ir.it)

17

IS

l'l

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Jl>i)

L'.">

Vent and Head Spra\Steam DrverSteam Outlet Flow Resirietoi

Strain Sepai ,ltoi s

RPY Stahili/.erKeeduatei Sparger

Shutdown Cooling OutletLow Pressure Flooder (l.PFL)

and Shutdown Cooling SpargerIligh Pressure Core l'looder

(HPCK) Sparger

HPCF Coupling1 o|.i (.luideKuel Assemlilies

< ,i ire Shroudt' i i > i( .onli ol Ki id

Core Plate

In ( .ore Instrument (.uule

T i l l i e s

( A m i n il R o d ( . H i d e T i l l i e s

C o l e 1 ) 11 f t_-1 e n l i a l 1 ' l e s s i i i e l i n e

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C o u t i o l Rod I ) i i \ e H o u s i n g s

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•

^M

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ÉJ11

Pi

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Figure A KB IABWR: Reactor pressure vessel and inlernals.

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1995-02-13

AEB2:i(3)

AEB2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Schedule

Description

BWR 90

NPP

BWR


ABB Atom, Sweden

Commercial

The BWR 90 is an advanced BWR plant design, based on the reliableoperation of BWR 75 design Forsmark 3 and Oskarshamn 3 units in Sweden.Modifications are mainly moderate and have been made to adapt to updatingtechnologies, modern safety requirements and to achieve cost savingsmeasures for simplified operation, testing and maintenance.

The core design is based on the advanced fuel assembly design, SVEA-100,which uses thinner rods (9.62 mm in diameter instead of 12.25 mm forSVEA-64 fuel assembly) and provides improved moderation and increasedheat transfer efficiency. The design margins for the fuel have increased, andsome of these margins have been utilized to raise the power levels, comparedwith predecessors. The advanced burnable absorber strategy, with axial andradial grading is maintained. In the BWR 90 plant core cooling is providedby 8 internal recirculation pumps, as in BWR 75 plants, although somewhatlarger in size. Also the prestressed concrete reactor containment is equippedwith filtered venting and an extensive redundancy and separation of safety-related systems has been installed.

The BWR 90 plant is currently ready for commercial use. No BWR 90 planthas been ordered yet. The BWR 90 was offered to Finland as a contender forthe fifth Finnish nuclear power plant that was cancelled last year.

ww victor/AEB2 ea

STUDSVK ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

AEB2:2(3)

BWR 90 Unit Data

Overall plantThermal powerNet electric power

Reactor plantAssembly geometryNumber of fuel elements per assemblyTotal uranium weight in first coreSpecific power of reactor coreTotal coolant flow rateFeedwater temperatureCoolant temperature at core outletReactor operating pressureAverage fuel burnup

Turbine plantTotal main steam throughput at full loadTurbo-generator powerNumber of turbine casings

3 300MW1 170 MW

4 x ( 5 x 5 )100126152KW/113 100kg/s215 °C286 °C70 bar47MWd/kg

1 775 kg/s1215MW1 HP + 3 LP

References LONNERBERG, B and PEDERSEN, T JBWR 90: An evolutionary ABWR plant for the next decadesASME/ISME Nuclear Engineering Conference, Vol. 2, ASME, 1993,p 633-638.

ww victor/AEB2 ea

STTDSYIK F-( O & SAFETY AB STl DSYIK i:S ^ !• AKB2:3.3.

Figure A Fit 2| i \VR l)D: KfiH

w,v VK !

VRIDS VIK ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

AEHl:i(3)

AEH1

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

CANDU9

NPP

HWR


AECL (Atomic Energy of Canada, Ltd), Canada

Commercial

The first CANDU NPP was taken into operation in 1962. Currently, thereare about 30 CANDU reactor units operating, or under construction aroundthe world. Four CANDU 6 units are in operation and five are under con-struction in three different countries. The design rating of the Mk I units is ofthe order of 650 MWe, while the Mk II has a net rated output of about800 MWe; increased ratings mean a lower special capital cost The powerincrease was achieved with only minor changes in design and equipment.

The CANDU 9 design follows the same evolutionary path. This reactor plantalso incorporates advanced features: reduced capital and generating costs,and a minimized construction schedule, achieved by a high level of standard-ization and modularization. The more extensive use of computers for safetysystems, the use of multiplexing and data highway technology reduceinstrumentation and control wiring. Self-checking is also used.

The use of slightly enriched uranium (1.2 % U-235) in CANDU 9 in place ofnatural uranium opens the opportunity to increase electrical output to therange of 1 050 MWe within the same operating limits as CANDU plants nowin service.

The four safety systems comprise the two diverse, dedicated reactor shut-down systems (gravity-drop absorbing rods and liquid neutron absorber).The emergency core cooling system uses ordinary water for injection.Improvements incorporated in the CANDU 9 emergency cor; coolingsystem (ECCS) include:

o Placement of all ECCS equipment, except the gas tanks, withinthe reactor building, will eliminate the need for a number ofisolation valves, thereby simplifying the ECCS system.

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1995-02-13

AEH1:2(3)

Schedule

o The use of one-way rupture discs to separate the heat transportsystem from the ECCS, thereby greatly simplifying the ECCSand reducing both the capital and maintainance costs.

No CANDU 9 plant has been ordered yet The detailed design is under waynow.

CANDU 9 Unit Data

Electrical output

Primary pressure

UO2 enrichment

Assembly

Number of fuel channels

Primary circuit configuration

Reactor pressure tubeMaterialDiameterHeight

Containment construction

1 050 MWe

80 bar

1.2%

37-elemcnt bundle

480

Loop

Zr-2.5 Nb0.1m6.0 m

Steel lined reinforcedconcrete

References FEHRENBLACK, P J and DULTON, REvolutionary of CANDU technology: a status reportAdvanced technologies for water cooled reactors.IAEA, Vienna, Austria, 1993, IAEA-TC-633.20.

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1995-02-13

AEH1:3(3)

Figure AEH1CANDU 9: Reactor building.

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1995-02-13

AT: 1(3)

AT Evolutionary Designs for Medium Power Plants

ATO Introduction

The medium power (600 MWe or less) range for most of these designs iscompatible with the present trend of reduced growth rates in electricityproduction in the developed countries and will more easily connect with localand limited electrical networks of developing ones.

Also, to meet the economic challenges posed by cheap oil and coal, somedesign teams have looked at two power levels: 600 MWe and 1 000 MWeranges (e.g. SBWR, AP-600 and SWR-1000).

Common to most of the evolutionary designs for medium power plants is asignificant departure from present-day designs. In this case, the design safetyprinciples addressed to different functions and systems are integrated intocomprehensive and balanced sets of measures at the NPP level.

The main features and objectives of the considered approach are thefollowing:

o Plant simplification, both at system and at component level,including construction and operational characteristics. Thisaims to save plant costs, along with standardization

o Additional safety margins on the "traditional" design basisaccidents (basically increased thermal margins for the fuel andcoolant inventory)

o Extensive application of inherent safety characteristics (e.g.temperature, density and size feedbacks) and passive safetyfeatures (e.g. gravity, natural circulation, evaporation,condensation e.a.)

o Slow accident progression rates (i.e. the reaching of criticallimits of response parameters is delayed under perturbations)

o

o

The capability to cope with any considered event for a pre-determined "grace period" without necessarily relying onhuman actions for the management of the accident

Improved reliability of safety functions, with particularreference to reducing human error importance and systemfailure occurrence rates

Protection against a large set of accidents including severe coredamage accidents with failure of the reactor vessel

ww victor/AT ea

STUDSVUC ECO & SAFETY AB STUDSVnOES-95/10

1995-02-13

AT:2(3)

o Reduction of the rate of initiating faults (e.g. deviations from

normal operation requiring reactor scram)

o Simplification of emergency preparedness requirements

o Adoption, as extensively as possible, of fully proven, state-of-

the-art technologies, without the exclusion of promisingtechnological advances

Considerable experience accumulated in construction and operation has ledpower-producing companies to renew their interest in medium power plants.

As noted above, medium power range is more open (due to the lower unitpower and decreased specific core power) for the implementation of thenatural processes in die safety systems. It is expected that reliance onpassive* safety features will lead to a better understanding by the generalpublic and recognition that a major improvement in public safety has beenachieved.

The use of passive systems and components offers a number of obviousadvantages, but may introduce new problems, such as:

o Different design methods that have not yet been definitelyassessed, including the tools to evaluate performance,capabilities and margins

o

o

o

More difficult thermohydraulic circuit-balancing, including thecharacterization of natural circulation loops (with smallpressure drops across the circuit) when operating together withactive (forced convection) systems

Augmented importance of phenomena such as thermal strati-fication and the presence of incondensable gases

Increased importance of heat losses and circuit layout to avoidconditions adverse to the initiation of natural circulation

Low water heads (especially in the long term), requiring largediameter pipes for water/steam flows and to avoid steambinding and small differential pressure heads to open checkvalves that normally respond to large reverse pressures

Difficulties in testing and maintenance during normal operationand in assessing availability for emergency operations

ww victor/AT ea

The word "passive" is perceived as negative by the general public but it ishard to find a relevant short expression: why not "natural"?

STUDSVK ECO & SAFETY AB STUDSVTK/ES-95/10

1995-02-13

AT:3(3)

Medium power evolutionary technology LWR designs are now under con-sideration in the USA, China, Great Britain, Japan, Russia and some othercountries. Only one mid-design represents HWR. It makes full use of proventechnology with relevant features resulting from the Canadian research anddevelopment program. Another one represents the continuation of theRussian RBMK line.

References

PEDERSEN,TJCurrent trends in safety philosophy and design goals for advanced nuclearpower plant designs.Proceedings ASME/TSME Nuclear Engineering Conference, Vol 2,1993ASME,p 539-544.

Safety aspects of Designs for Future Light Water Reactors (EvolutionaryReactors).IAEA, Vienna, 1993, July, IAEA-TECDOC-712.

International Nuclear Congress - Atom for EnergyProceedings ENS-94, Vol 1,1994, October 2-6, Lyon, France.

ww vicior/AT ea


1995-02-13

ATPl:i(3)

ATP1

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

WER-500/600

NPP

PWR

Medium power/600 MWe

EDO Hydropress, RF

Detailed design

The WER-500/600 reactdeveloped on the basis of well-established W E R practice. Approaches areclose to the world tendency for medium-power loop-type advanced PWRplant designs:

o

o

o

o

o

o

o

o

o

o

Reduction in the unit power

Reduction in specific core power density

Simplification of the structural scheme of the reactor plant

Use of proven technical solutions

Four 60 bar accumulators and four 36 bar accumulators

Coolant natural convection for residual heat removal

Primary circuit depressurization system opening to pressuredecrease down to 3 - 5 bar

Compensation for water losses from the steam generator bypassive systems

External water cooling for the steel containment

4 diesel generators (2 large and 2 small)

In the design the four-loop reactor with horizontal steam generators is used.The prototype reactor is the VVER-1000 reactor design. The core powerdensity is about 65.4 kW/l and the reactor pressure vessel has been maderather large (the same as the VVER-1000) in order to guarantee large waterinventory.

The core with absorber rods inserted is subcritical at the initial moment ofthe cycle for a coolant temperature less than 100°C even in the event of

ww victor/ATP 1 ea


1995-02-13

ATP1:2(3)

Schedule

berated water substitution for pure condensate. The efficiency of fuel isincreased when compared to operating NPPs equipped with WER-1000 by20 - 25 %. The service life of the major equipment is extended up to50 years.

The containment has a cylindrical arrangement: the inner shell is leaktightmade of steel, the outer shell of reinforced concrete. Filter vents from theleaktight shell and corium trap installation are envisaged.

The NPP is designed to survive blackout under normal operating conditions,operating events and design basis accidents for not less than 24 hours.

The design envisages NPP construction in different climatic conditions and inseismically dangerous areas with an earthquake intensity of 8 points as perMSK-64.

For regions with an earthquake intensity of 8-10 points the application oflow-frequency seismic isolating devices reducing the seismic loads by 10-30times is envisaged.

No WER-500/600 plant has been ordered yet. Some testing programs are inprogress to support the final design and licensing process.

VVER-500/600 Unit Data

Rated thermal power

Primary coolant pressure

Primary coolant average temperaturs

Pressure of produced saturated steam

Average core power density

Average bumup

Average makeup fuel enrichment

Reactor vessel neutron fluence (for 50 years)

Containment gross volume

1800MW

157 bar

320 °C

70.6 bar

65.4 kW/l

39.6 MWday/kg

3.45 % (mass)

2.51019n/cm2

60 000 m3

References VOZNESENSKY, V, KUHTEVITSH, I and ROGOV, MAdvanced NPP WER-500 with passive safety systems.Proceedings International Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 25-29, 1992, Tokyo, Japan, v 1, p 44.

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1995-02-13

ATP1:3(3)

1. Reactor2. Main coolant pump3. Steam generator4. Pressuriser5. Hydroaccumulator6. Low pressure tank7. Tank of Chemicals foriodine precipitation8. Spray header9. Special heat-accumulatingtank10. Heat exchanger of passiveheat removal system (PHRS)

11. Connecting valves12. Safety bypass valves

13. Regenerative heatexchanger14. Letdown heat exchanger15. Ion exchanger16. Pressure relief system17. Barboter-degasator18. Spent fuel storage pool(SFSP)19. Pipeline between SFSPand accidental heat removalbasin

20. Coolers of containmentheat removal system21. Hermetic steel shell22. Regenerative heaiexchanger23. SG water cleanup system

24. SG water cleanup systempump25. SFSP cooling system pump26. SFSP cooling system heatexchanger27. Primary circuit chargingpump28. Primary water dischargecooler29. Makeup surge tank30 Intermediate cooingwater pump31. Intermediate water cooler32. Dry cooling lowc33. Essential service water

Figure ATP1VVER-500/600: Safety and normal operation systems.

ww victor/ATP 1 ca

STUDSVnC ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

ATP2:i(4)

ATP2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

AP-600 (Advanced Passive)

NPP

PWR


Westinghouse, US

Detailed design

The AP-600 is a plant desigiBechtel, Avondale Industries, Burns & Roe and others. The reactor coolantsystem comprises the RPV, two vertical SGs, four RCPs, pressurizer andlinking pipework. A two loop Westinghouse PWR arrangement wasmodified to have conservative safety margins and permit simplification ofmany supporting subsystems. The AP-600 features incorporate an improvedRCS configuration utilizing canned motor reactor coolant pumps directlycoupled to the steam generator bottom. This eliminates the "cross-over leg"from the RCS, lowering the overall system flow resistance, and improvingthe performance with respect to small break loss-of-coolant accidents. Thisgreatly simplifies the RCS loop support configuration.

The thermal power rating of the reactor core is 1 812 MW. The fuelassembly design is basically the same as in other Westinghouse PWRs, butthe power density has been reduced by approximately 30 %, the linear heatrating has been reduced correspondingly. Combined with the introduction ofa radial neutron reflector, this has reduced the enrichment needs for the fuel.The low power density core, the radial reflector and a somewhat largerdistance between the core and the RPV wall have resulted in a significantreduction in neutron fluence. The reactor internals are essentially of a con-ventional design.

The large reactor vessel and the increased core coolant flow cross sectionreduced the hydraulic resistance of the RCS to the level at which 17 - 25 %power can be removed by natural convection at nominal coolant heating. Theincreased water inventory resulting from increased RPV and pressurizervolumes improves the reactor safety ability.

The key safety features of the AP-600 reactor plant arc:

o Small leaks up to 6" diameter can be handled without RCSdepressurization

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1995-02-13

ATP2:2(4)

Schedule

References

o For large pipe breaks the peak clad temperature does notexceed 980 °C

o Containment cooling is accomplished by a combination ofwetting and the natural circulation of air through gaps betweenthe shield building and the steel containment

o Passive systems are expected to perform their safety functionsfor up to 72 hours after the initiation event, independent ofoperator action and off-side power

This approach provides an opportunity for higher availability, NPP simpli-fication, reduction of costs and a shorter construction period. AlsoWestinghouse began working with Japanese utilities and EdF (France) toinvestigate the feasibility of applying the AP-600 passive concepts to a threeloop 1 000 MWe reactor plant, known now as AP-1000.

The AP-600 development programme to obtain a design certification fromthe US NRC by 1996 is well underway. Some testing programs are inprogress to support final design and licensing processes. No AP-600 planthas been ordered yet, but Westinghouse, and its associates in the design teamhave been awarded a FOAKE contract with US utilities and US DOE (1996Sept).

AP-600 Unit Data

Gross electric power 630 MWe

Number of cold pipclegs 4

Number of hot pipelegs 2

Number of fuel assemblies 145

Type of fuel assembly 17 x 17

Number of fuel elements 38 280

Maximum coolant tcr-perature 324.4 °C

Average linear heat rating of fuel elements 12.6 kW/m

Average core power density 73.9 MW/m3

Reactor fuel loading 61.021

Reactor vessel inner diameter 399 cm

Neutron fluence for reactor vessel (for 60 years) 2 • 1019 n/cm2

Safety aspects of designs for future LWRs (Evolutionary reactors)IAEA-TECDOC-712, July 1993, Vienna, Austria.

The Westinghouse AP-600Nuclear News, 1992, September, p 76-77.

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1995-02-13

ATP2:3(4)

Figure ATP2AP-600: Reactor plant layout.

ww vic(or/ATP2 ea

The Westinghouse AP600

1 Fuel handling area2 Concrete shield building3 Statt containment4 Passive containnwnt cooing water tank5 Passive containment cooing air baffles6 Passive containment cooing air Mats7 Equipment hatches (2)*8 Personnel hatches (2)9 Core makeup tanks (2)

10 Steam generators (2)11 Reactor coolant pumps (4)12 Integrated head package13 Reactor vessel

\ Pressuruer15 Depressurization valve module location16 Passive residuai heat removal heat exchangers17 RefueHing water storage tank ,18 Technical support centre•19 Main control room "'"1 integrated protection cabinets j

I High pressure feedwater heaters! Feedwater pumps

Deaeratof' Low pressure feedwater heaters

Turbine generator


1995-02-13

ATP3:i(3)

ATP3

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

MS-600 (Mitsubishi Simplified)

NPP

PWR

Medium power/630 MWc

Mitsubishi, Japan

Basic design

The MS-600 reactor plant has aenhanced safety, reliability and cost reduction.

The thermal power of the reactor is 1 825 M W L The core consists of17 x 17 array assemblies and the volume power density has been lowered by30 % of that of present-day large power PWR plams. Neutron economy andfuel cycle costs are improved by means of a steel radial neutron reflector.Core instrumentation is top-mounted, in turn, this simplifies the containmentdesign and facilitates inspection and maintenance.

Two horizontal steam generators with increased water inventory are used inMS-600. Previously, it was a special feature of Russian W E R designs. Thiswould eliminate a lot of traditional vertical SG problems due to crud depo-sitions and subsequent stress corrosion cracking. The reactor coolant pumpsare of a more efficient type, and with improved shaft sealings.The primarysystem, the spent fuel storage and the water tanks for safety injection arelocated inside a spherical steel containment A secondary containment wasmade up by a concrete filled steel. The space between the primary andsecondary containments is vented to the enviroiiment via a passive filteringsystem.

The safety systems are based upon current technology with application ofnew materials and have been simplified to improve reliability and operation.The MS-600 plant design uses active systems for the minor accidents (smallLOCAs, transients without flooding the cointainnient) and passive systemsfor the serious accidents.

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ATP3:2(3)

Schedule

The active sytems includes pumps for safety injection, auxiliary feedwatersupply and small diesel generators. Passive safety functions are achieved byan automatic depressurization system, pressure and gravity-drivenhydroaccumulators.

A detailed design and design artification test program is scheduled forcompletion by 1997. An expanded development effort, involving cooperationwith utilities in Japan, is planned for the near future, aiming at the design of a1 200 MWc NPP based on the MS-600 principles. This programme has beengiven a high priority with respect to development of the next PWRgeneration in Japan.

MS-600 Unit Data

References

Electrical output

Average linear heat rating of fuel elements

Fuel assembliesTypeNumber

Turbine

Reactor coolant systemNumber of loopsOperating pressure

TemperatureReactor outletReactor inlet

Steam generatorsNumberType

Steam pressure

Reactor coolant pumpsNumber

630 MWe

15.1 kW/m

17x17145

TC4F40

2154 bar

325.0 °C290.6 °C

2Horizontal, U-Tube type

57 bar

2

Technical Information on Design Features of MS-600Nuclear Power Engineering Corporation, Tokyo.Review of Advanced Light Water Reactor Design Approaches.10-13 May, 1994, Moscow, RF, IAEA-TC-879 .

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1995-02-13

ATP3:3(3)

Secondary depressurisation valves

Condensate storage tank

Gravity injection pit Primary depressurisation valves

Advanced accumulators

| I Horizontal steamgenerators

Pressunzer i

Auxiliary feedwaterpumps

Charging/SI pumps

Figure ATP3MS-600: Configuration of hybrid safety systems.

ww victor/ATP? ca


1995-02-13

ATP4:i(3)

ATP4

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Schedule

Description

AC-600

NPP

PWR


China National Nuclear Corporation, China

Basic design

The AC-600 design is expected to improve the economy and safety ofcurrent standard PWR designs by using system simplification, passive meansand modular construction. Many features of the AC-600 reactor plant aresimilar to the well-known Westinghouse AP-600 design (see ATP2).

AC-600 has an increased margin of operation because of a decreased corepower density. Increased natural circulation flow rate in RCS is an importantfactor of the AC-600 concept The increased volume of reactor pressurevessel and pressurizer improves the reactor ability for self-control and makesthe black-out emergency situation milder. The AC-600 concept eliminatesthe high pressure safety injection pumps, utilizes the full pressure core make-up tanks and larger hydroaccumulators.

AC-600 will become a major type of reactor for the next generation600 MWe nuclear power plants in China. The conceptual design iscompleted and is ready for detailed design development.

ww victor/ATP4 ea

JTUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

AC-600 Unit Data

Rated thermal power

Fuel assemblyArrayNumber of fuel rodsOutside diameter

Reactor coreNumber of fuel assembliesTotal weight of UO2

Reactor coolant systemNominal pressureVessel outlet/inlet temperatureAverage linear ratingAverage core power densityNeutron fluence for 60 years

Steam generatorTypeNumberHeat transfer surfaceWater volume of 2-nd side

ContainmentConfigurationGross volume

References MIN YUAN-YOU e.a.

A1T4:2(3)

1936MW

17x1738 2809.5 mm

14566.81

15.8 bar327/293 °C134.2 W/cm77kW/l2 • 1019 n/cm2

Vertical, U-tube25 430 m2

169 m3

Double (steel/concrete)50 000 m3

Basic Information of Design features of AC-600 Advanced Reactor Plant.Review of ALWR Design Approaches.10-13 May, 1994, Moscow, RF, IAEA-TC-879.

ww victor/ATP4 ea

STUDS VIK ECO & SAFETY AB

INLET PLENUM

STUDSVIK/ES-95/10

1995-02-13

CONTAINMENT COOUNG

HEATED AIR DISCHARGE

ATP4:3(3)

PCCS WATER STORAGE TANK

CONTAINMENT COOUNGAIR INLET

CONCRETE

BASEMAT

STEEL CONTAINMENT VESSEL

- AIR BAFFLE

Figure ATP4AC-600: Containment cooling.

CONCRETE

SHIELD

ww victor/ATP4 ca

STUDSVDC ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

ATBl:i(3)

ATB1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SBWR (Simplified Boiling Water Reactor)

NPP

BWR


General Electric, US

Detailed design

The design of the SBWR is the common reinvolving organizations in Europe, Asia and North America. This BWR isdesigned to meet passive ALWR requirements, including no dependence onoperator action for three days after a core damaging event.

The SBWR key features are the following:

o An integral dryer-separator simplifies refuelling

o A large reactor vessel water inventory increases margins and

enhances safety

o Standard failure-resistant fuel

o A low power density core

o Recirculan'on system eliminated (natural circulation)

o Fine-motion control rod driver (CRD) for easy operation andmaintenance

The power rating of the reactor is 1 800 M W L The nuclear fuel is similar tothat used in General Electric BWR designs, but due to the use of the naturalcirculation, the core height is reduced by about 30 %. The volumetric corepower density decreased up to about 36.6 kW/l. Also in the SBWR the RPVis 2.6 m higher than that required for the 1 356 MWe ABWR (see AEB1).The natural circulation, compared with the forced circulation, means loweredfuel cycle costs, a reduced number of transients and increased thermalmargins.

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1995-02-13

ATB1:2(3)

Schedule

The safety features of the SBWR are a gravity-driven core cooling systemthat will keep the core covered and cooled in the event of a LOCA, anisolation condenser located in an elevated water tank, which allows forresidual heat removal by natural circulation, and a containment of thepressure suppression type with a passive containment cooling system(PCCS). The isolation condensers and the PCCS passively transfer decayheat to the atmosphere. In a LOCA, the reactor vessel is depressurized bymeans of depressurization valves, the gravity driven cooling system floodsthe reactor and the isolation condenser removes the decay heat There is nocontainment flooding for most LOCA events. The flooding of the lowerdrywell occurs mainly in response to severe accidents. There is a sufficientinventory in PCCS pools to handle at least 72 hours of decay heat removal.

The design team has also looked at a 1000 MWs power level for SBWR. Indie area of standardization and overall plant economics, the SBWR 1000 islikely to satisfy the utility needs for a simplified and economic plant in thecoming decade (see ATB4).

No SBWR plant has been ordered yet An SBWR development programmeFOAKE to obtain FDA by mid 1994 and FOAKE by the end of 1996 is wellunderway. Some large scale and full scale tests are still in progress.

SBWR Unit Data

Electric power

Primary pressure

Power density

Reactor pressure vesselDiameterHeightMaterial

Refuelling time

Reference fuel cycle length

Plant design life

600 MWe

72 bar

36.6 kW/l

6 m24.6 mSteel

- 1 5 days

24 months

60 years

References Safety Aspects of Designs for Future Light Water Reactors (Evolutionaryreactors)Vienna, Austria, 1993, IAEA-TECDOC-712

Simplifying the BWRAtom No. 430,1993, Sept/Oct

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1W5-O2-13

ATB1:3(3)

Figur ATB1SBWR: Reactor vessel and internals.

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ATB2:i(3)

ATB2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

HSBWR (Hitachi Small Boiling Water Reactor)

NPP

BWR


Hitachi, Japan

Basic design

The HSBWR is a proposed natural circulation BWR plant, with a thennalpower of 1 800 MW. The core power density is about 70 % of that ofcurrent BWRs because of natural circulation.

The main features of the HSBWR design are as follows:

o Short fuel assembly (an active fuel length of 3.1 m) is selectedto improve seismic resistance of the core.

o A riser, 9 m in height, above the core and elimination of steamseparators will increase the rate of natural circulation.

o A low power density of 34.2 k.W/1 allows for the extension ofthe operating cycle by up to 2 ycrs .

o Simple reactor components and systems without forcedrecirculation systems and active emergency core coolingsystems.

o No core uncovery during any LOCA events in the reactorpressure vessel; coolant injected by the actuation of the steam-driven reactor core isolation cooling system, automaticdepressurization system and accumulators.

o A pool of water surrounds the metal primary containment andheat rejection from the containment to the pool by conductionthrough the containment wall removes decay heat for 72 hourswithout intervention.

o Reactor pressure vessel bottom flooding by spillover ofemergency core cooling system water.

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1995-02-13

ATB2:2(3)

Schedule

o Depressurization by an automatic depressurization system andborated water injection by the accumulator makeup plantshutdown during the anticipated transient without scram(ATWS).

o Decreased investment costs by reducing the construction periodup to 32 months.

Conceptual design complete. The HSBWR is apparently still in thedevelopment stage. No HSBWR plant has been ordered yet.

HSBWR Unit Data

Electrical output

Rated pressure

Core flow

Core length (total/active)



Core equivalent diameter

Power density

Uranium enrichment

Average burnup exposurs

Recirculation system

RPV diameter

RPV height

Steam separator

Turbine type

Size of reactor building

600 MWe

70 bar

178 000t/h

3.7/3.1 m

708 ( 8 x 8 array)

169

4.7 m

34.2 kW/l

3.6%

39MWd/kg

Natural circulation

6.3 m

25 m

None free surface separator

TC 2F 40

47 m x 47 m x 47 m

References Small and Medium Reactors, Technical SupplementNuclear Energy Agency, OECD, Paris, France, 1991

Safety aspects of designs for future light water reactors (evolutionaryreactors)IAEA-TECDOC-712, IAEA, Vienna, 1993

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1995-02-13

ATB2:3(3)

ACCUMULATOR

STANDBY LIQUID REACTORCONTROL

STEEL PRIMARYP V E I 1 E L E CONTAINMENT VESSEL

RESIDUAL HEATREMOVAL SYSTEM

OUTER POOL

REHEATER

CORE SUPPRESSION

GENERATOR

CONDENSER

2

POOL RCIC(St«am Driven) RESIDUAL HEAT

REMOVAL SYSTEM

FEEDWATER PUMP

Figure ATB2HSBWR: Schematic.

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ATB3:i(3)

ATB3

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

TOSBWR-900P (Toshiba Simplified BWR)

NPP

BWR


Toshiba, Japan

Basic design

The TOSBWR-900P re«ctor plant is a propcirculation BWR. The main number of features for the TOSBWR-900Pdesign arc similar to those of the GE SBWR. The primary design objectivesare: to simplify the plant systems, to decrease investment costs, to reduce theconstruction period and use passive safety measures to improve the safety.

The main features of the TOSBWR-900P include:

o The reduced size of the reactor vessel due to shorter fuelassemblies and the elimination of steam separators and driers(steam drums)

o The use of hydro accumulators for water injections and anautomatic depressuriza'ion system

Decay heat removal provided by a gravity-driven coolingsystem and passive containment spray

Top entry gravity-driven control rods made possible by theelimination of components above the core. This design pro-vided the elimination of all penetrations in the RPV bottom

Steel containment wall which can transfer heat from the floodedcontainment to an external pool which uses sea water as a heatsink

Isolation condensers located at an elevated position in thesuppression pool in the wetwell region in the primarycontainment

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ATB3.*2(3)

Schedule Conceptual design complete, ready for detailed design development NoTOSBWR-900P plant has been ordered yet.

TOSB WR-900P Unit Data

Thermal power

Electrical power

Type of fuel assemblies


Effective core size (diameter/height)

Power density

Size of pressure vessel (diameter/height)

Operating pressure

900 MWt

310 MWe

8 x 8 array

388

3.4 m/2.5 m

40kW/l

4.7 m/17 m

72.1 bar

References Safety aspects of designs for future light water reactors (Evolutionaryreactors)Vienna, Austria, 1993, IAEA-TECDOC-712.

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ATB3:3(3)

DRYWELL

GRAVITY-DRIVENCOOUNG SYSTEM

SECONDARYCONTAINMENT

VESSEL

STEEL WALL TOTRANSFER DECAY

HEAT TO SEAWATER

HEATEDSEAWATER

OUT

COLDSEAWATER

IN

POOL WATERDISCHARGE

VALVE

WETWELL

VENT PIPE

LOW-PRESSUREACCUMULATOR

ISOLATIONCONOENSER

STEEL WALL

CONTROL RODDRIVE

HIGH-PRESSUREACCUMULATOR

REACTORCORE

Figure ATB3TOSBWR-900P: Reactor building.

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1995-02-i:

ATB4:i(2)

ATB4

Format

Title

Application of theReactor

Reactor type

Power output

Organization (name)

Development status

Description

Description

1000 Natural Circulation BWR

NPP

BWR



Studies

The 1000 Natural Circulation E

Schedule

the 600 MWe power. The key differences are the power output andassociated adjustment of equipment sizing. Both plants use similar passivesafety features and a common approach for the reactor building arrangementand containment

The riser height and separator standpipe lengths were chosen to minimizeRPV height and refueling time impacts when scaling up to 1 000 MWeoutput. Compared to SBWR, the 1 000 MWe reactor has a shorter riserabove the core and longer separator standpipes. These differences permit ashorter span from normal water level to the second level isolation setpointfor the 1 000 MWe reactor, since the longer standpipes increase the in-vesselwater inventory. The effective stack height which determines the naturalcirculation flow driving force is 0.5 m greater for the 1 000 MWe reactor. Byproviding a greater in-vessel water inventory while maintaining the effectivestack height, the overall height of the two reactors differs by less than ameter.

A higher steam flow is accomodated by increasing the number of steam linesfrom two in SBWR to four in the 1 000 MWe design. However, the steamline diameter is reduced from 1 000 mm to 600 mm. The number of safetyrelief valves is increased from 8 to 14 and the additional ones are placed onthe additional two main steam lines. The total number of depressurizau'onvalves is increased from 6 to 8 (the additional ones are installed on each ofthe two additional stean: 'ines).

It was shown by calculations that from the viewpoints of natural circulationapplication and passive safety features, the SBWR design can be scaled up toa 1 000 MWe capacity.

No 1 000 MWe Natural Circulation BWR has yet been ordered.

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ATB4:2(2)

References

1000 Natural Circulation BWR Unit Data

Power 1 000 MWe

Reactor pressure vessel

Diameter 7.1 mHeight 25.2 mMaterial Steel

Cycle length (effective full power days) 357

Maximum linear heat generation rate 39.3 kW/m

HSU, L C and FENNERN, L E

1000 MWe Natural Circulation BWR TechnologyProceedings International Conference on Design and Safety of AdvancedNuclear Power Plants. Vol. II, October 25-29,1992, Tokyo

Figure ATB41000 Natural Circulation BWR: Gravity driven cooling system (GDCS).

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ATB5:i(3)

ATB5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SWR1000

NPP

BWR


Siemens-KWU, Germany

Basic design

The SWR-1000 is an advarelectrical output of 1 000 MWe. The main features of SWR-1000 safetyconcepts are:

o A higher degree of safety is achieved through the introductionof passive systems for accident prevention and control

o In the event of an accident only after several days ofintervention by operating personnel is necessary

o A large RPV with a low core power density, large reactorcoolant inventory and an increase in the water inventoriesstored inside and outsiiie containment

The core design is based on an enlarged Type 12 x 12-16Q fuel assembly,which was designed by adding two additional fuel rods to each axis of a10 x 10 array fuel assembly. This assembly is characterized by a loweraverage rod linear power density of 108 W/cm. The number of controlassemblies in the core has been reduced by approximately 40 % to 157, and anumber of detector assemblies by some 60 % to 18.

The passive safety systems of SWR-1000 are the following:

o Emergency condenser, removes decay heat from the reactor atthe RPV drops, requiring neither electric power nor activationby the I&S system

o Containment cooling condensers remove heat due to steamformation inside the containn^nt

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ATB5:2(3)

Schedule

o A gravity-driven core flooding system floods the reactor bygravity flow during RPV and pressure drops

The containment is a pressure suppression-type design, provided with agrace period for the initial 7 days (hypothetical loss of coolant core melt forwhich a 100 % Zi-H^D reduction is postulated) and an inert atmosphere forprevention.

No SWR 1000 plant has been ordered yet

SWR 1000 Unit Data

Overall plantThermal outputNet electric output

Reactor coreNumber of fuel assembliesTotal uranium weightActive core heightAverage power densityDischarge bumupAverage enrichmentReload fuelCoolant flow rate

Reactor pressure vesselInside heightInside diameterDesign pressureNumber of rccirculation pumps

TurbineInlet flowInlet pressureNumberSpeedNumber of HP casingsNumber of LP casings

Plant operationSpent fuel storage capacityPlant design lifePlant construction period

2 778 MWt977 MWe

648127 t2.8 m48kW/l65 GWd/t4.95 wt %13t/a12 000 kg/s

22.8 m7 m88 bar6

1 475 kg67 bar13 000rpm13

40 years60 years48 months

References SWR 1000 Technology of the future. Medium capacity boiling WaterReactor With Passive Safety Features.Siemens AG - Bereich Energieerzeugung (KWU), Germany Aug 1994.

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1995-02-13

Containment cooing«ond«nMt (1 et 4)

; flenaor water clean up -ysiem(no salely syilo

] Core calcner

k, 'n.'.v,!1 \o w h ' ti' t i E S Sr~"Re8k)ual rieal removal syslem

{no 3alei '

~—-:—i cold water• • • I hot waterSiflFSFKl steam

Fijjure ATB5SWR 10(10: Contiiinnient building.

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1995-02-13

ATGl:i(3)

ATG1

Format

Title

Application of theReactor

Reactor type

Power output

Organization (name)

Development status

Description

Description

MKER-800

NPP

LWGR


RDIPE.RF

Studies

With a net electrical powc

Schedule

represents the latest version of the post Chernobyl RBMK line. It makes useof the proven technology of current RBMK plants. In this graphite channelreactor with water cooling, the lessons learnt from operating RBMKs andfrom the Chernobyl disaster have been incorporated.

MKER-800 reactor plant safety and reliability improvements include:

o Reasonable reactivity characteristics of the core

o Natural circulation providing heat removal up to 600 MWe

with an additional application of the jet pumps at nominalpower

o Absence of any valves in the primary circuit

o Increased efficiency of the emergency shut-down system

o Decreased linear specific power to 28 kW/m

o Containment with improved seismic stability

o Passive cool-down system located inside the containment

No MKER-800 plant has been ordered yet Conceptual MKER-800 designcomplete; ready for detailed design development

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ATG1:2(3)

MKER-800 Unit Data

Reactor thermal power 2 450 MWt

Electric power 860 MWe

Core linear power 28 kW/m

Steam rate 4 600 t/h

Average burnup fuel 28 MWd/kg

Steam pressure 70 bar

Fuel enrichment 2.4 %

Temperature of the feeding water 200 °C

Thermal load 250 Gcal/h

Electric power in the regime without heating 860 MWe

Electricity consumption for needs of the plant 4.5 %

Power unit efficiency/net 33.5 %

References ADAMOV.EOGraphite-channel family: Status and perspectives.Book of Abstracts. Nuclear Power & Industry International Conference,27 June - 1 July, 1994, Obninsk, RF.

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1995-02-13

ATG1:3(3)

Figure ATGlMKER-800: Reactor building.

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1995-02-13

ATHl:i(4)

ATH1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

CANDU3

NPP

HWR

Medium power

AECL (Atomic

Detailed design

With a net elecWith a net electrical power output of 450 MW the CANDU 3 represents thesmallest version of the CANDU-reactor line. It makes full use of the proventechnology of the established 665 MW mid-size CANDU 6 being updatedwith relevant features resulting from the ongoing Canadian research anddevelopment programme.

The CANDU 3 is fuelled by natural uranium and permits on-power refuellingas well as low excess reactivity. The fuel bundles are situated on Zr-2.5Nbhorizontal pressure tubes being surrounded by a Zircaloy calandria tube withCO2 as the thermal isolator in the annulus. Except for the pressure tubes allother components of the reactor assembly operate under low temperatureand low stress. The fuel bundles are cooled by pressurized heavy water andthe secondary heat transport system consists of one loop with two steamgenerators and two pumps. The arrangement allows for natural coolantcirculation in the event of loss of power in the heat transport pumps.

The additional cooling circuit for the heavy water moderator can also beused for decay heat removal in the unlikely event of LOCA with failure ofthe light water emergency core cooling system.

The main design goals aJm at:

0

0

Enhanced safety and a high lifetime capacity factor

Low occupational exposure (50 per cent reduction fromCANDU 6 to the reference design)

Maximization of the components/plant life (up to about100 years)

Easy replacement of any component

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1995-02-13

ATH1:2(4)

Schedule

o Short and manageable construction schedule (30 months)

CANDU 3 incorporates many safety enhancements, including:

o Simplified and increased containment design pressureeliminating the need for dedicated pressurr-suppression systems

o Increased pressurizer size and reduce ;eed condenser size tomaintain the heat transport syster • ^ntory over a wider rangeof upset conditions

o Improved moderator cir ••:'• »aon - to increase cooling capabilityfollowing a LOCA ar. oss of emergency coolant injection

o Improved man-' - . ine communication, simplifying centraloperations anc .iger operator action times (new control room)

o Severe accident mitigation

The CANDU 3 plant detailed design is under way and is scheduled forcompletion in 1996. No CANDU 3 plant has been ordered yet

CANDU 3 Unit Data

Power capacity

Number of fuel channels

FuelFuelForm

Length of bundleOutside diameterBundle weightBundles per fuel channel

Heat transport systemNumber of steam generatorsSteam generator typeNumber of heat transport pumpsHeat transport pump type

Reactor outlet header pressureReactor outlet temperatureReactor coolant flowSteam temperature (nominal)Steam pressure

1 439 MWt

232

Natural UO2Fuel bundle assemblyof 37 elements495 mm102.4 mm23.5 kg (18.4 kg U)12

2Vertical U-tube2Vertical, centrifugal,single section99 bar310°C5.3 Mg/s260 °C46 bar

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ATH1:3(4)

References Small and medium reactors, v 2 Techiucal supplement,Nuclear energy agency, OECD, Paris, France, 1991.

The AECL Candu 3 reactor.Nuclear News, September, 1992, p 80-81.

DAMPER OPENIN ACCIOENT

CONTAINMENTSTRUCTURE

DAMPERSCICSE FOR BUILDINGACCOENT A I R COOLERS

AIR FLOW FORVAULT C COL IN CANDMYOPOOENOlSPERSAi(IN ACCIDENT)

PENETRATIONS(CABINS PIPES)

NOTE VENTlllTiON SYSTEM »NO CUD VAPOUR RECOVERYARE NORMAL OPERATING SYSTEMS AND NOT PARTOF THE CONTAINMENT SYSTEM

VENTILATIONAIR SUPP'- i

Figure ATH1CANDU 3: Containment system.

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1995-02-13

ATH1:4(4)

OCTICTO» UKIT

»UN TM

rue i CHANMEI AsscuairIMCLUONC CALWKWK TUBECAUUKDIU IV6£ EXTENSIONA«0 €N0 »TTTINGl

UOUID MJCCTIO»SMUTDOWS L/NIT

Figure ATH1CANDU 3: Calandria/Reactor Assembly.

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1995-02-13

AI: 1(2)

AI Innovative Designs

AIO Introduction

The designers of these LWRs (600 MWe or less) concentrate their effortsprimarily towards achieving a reliable termination of the fission reaction inaccident situations using passive means and inherent reactor properties andtowards assurance of reliable passive cooling in the shut down reactor. Someinnovative designs incorporate significant departures from present-daydesigns in reactivity control, reactor coolant system configuration and othersystems.

If the reactor vessel integrity is not lost, a reliable way of terminating thefission reaction is core flooding with high concentrated boron water. For amore reliable core flooding under accidental conditions different methods areproposed in Sweden, Japan, Russia, Italy and the USA:

o Integrated arrangement of the primary circuit in a reactor vesselof prestressed concrete;

o Integrated arrangement of the primary circuit with a reactorvessel provided with a guard vessel;

o Accommodation of the whole primary circuit equipment in asystem of boxes of prestressed concrete, filled with "cold"water.

For reliable and rapid core flooding with "cold" water with a high boron con-centration, the PIUS-principle was developed in Sweden. This concept isused as a basis for some later designs in Japan both for innovative andcurrent PWR (PSSS - passive safety shutdown system) and in the USA forBWRs and steam cooled fast reactors.

The grace period is practically inversely proportional to the nominal thermalpower and directly proportional to the water inventory in the passive cool-down systems. To increase the self-protection time (grace period) in thePlUS-principle concepts RPVs with a water inventory of 0.3 - 1.5 m3/MWtare used, which permits a grace penod from 1 up to 7 days, respectively. Thedecay heat is removed by natural circulation.

Since the innovative designs deviate more than evolutionary ones frompresent-day technology, a prototype or demonstration plant is necessary fortesting and validation.

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AI:2(2)

References

o

GAGARINSKI, A Yu et alAdvanced LWRs: analysis of new approaches and ideas.Nuclear Society International, Moscow, RF, 1993.

'XÄSBERG.CWetalProposed and Existing Passive and inherent safety structures,systems and components for ALWRs.Report ORNL-6554,1989.

PEDERSEN.TJCurrent Trends in Safety Philosophy and Design Goals forAdvanced Nuclear Power Plants Designs.Proceedings ASME/ISME Nuclear Engineering Conference,Vol 2,1993, ASME, p 539-433.

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AIP1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

PIUS-600 (Process Inherent Ultimate Safety)

NPP

PWR


ABB Atom, Sweden

Basic design

The thermal power of the PIUS-600 PWR tyicore consists of 18 x 18 array fuel assemblies, with a reduced height to limitthe hydraulic pressure drop across the core during operation. The core ishoused at the bottom of a double-walled flow guide structure which is pro-vided with a wet metallic thermal insulation of the outer surface. Thisstructure is installed in a pool of highly borated water. The reactor pool,enclosed in a prestressed concrete reactor vessel (PCRV). The main designparameters (the core power density, reactor pressure and temperature) arered'/* vd in comparison to current PWRs.

<o or SGs and the four reactor coolant pumps are located outside the• *V. The PCRV is installed inside a primary containment of the pressure-

i- <}' :ssion type which, in turn, is largely enclosed in a reactor building.

I / J ig normal operation, and during minor disturbances, normal control1 .terns keep the plant in operation. With drastic deviations from nominalonditions safety systems will try to restore acceptable conditions or initiate/actor shutdowns. PIUS is provided with a scram valve system whichseparates the borated water pool from the primary loop. If the scram systemvere to fail, a self-protective mechanism would take over, the pressure

oalance in the density locks would be disrupted and pool water would enterle primary loop.

The main differences compared with traditional PWR designs are:

o The PCRV

o

o

The use of lower and upper density locks towards the PCR

The siphon breakers and wet thermal insulation

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AIP1:2(4)

Schedule

o The implementation of a passive residual heat removal systemwith a protected period of 7 days

o Reactivity control without control rods.

No PIUS plant has been ordered. Significant verification testing and designwork remain to be done, before a commercial application. This design willrequire a prototype demonstration before it can be artified.

PIUS-600 Unit Data

Electric power

Fuel assemblyNumber of fuel rodsCladding materialMean power density in fuel

600 MWe

312Zircaloy424.8 kW/kg

Reactor coreNumber of fuel assemblies 213Active height 2.50 mEquivalent diameter 3.75 mEnrichment (first core/reload) 2.0%/3.5%Average fuel bumup (equilibrium core) 45.5 MWd/kgTotal weight of U 801Volume core power density 72 kW/l

Reactor coolant system (operation conditions)Pressure at vessel inletPressure at vessel outletTemperature at pressure inletTemperature at vessel outlet

Prestressed Concrete Reactor VesselOverall height (without upper part)Overall widthInside diameterTotal weightDesign pressure

Reactor coolant pump type

Steam generator type

95 bar93 bar260 °C289.3 °C

44m26.8 m12.2 m63 0001105 bar

Variable speed, wet

Vertical once through

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AIP1:3(4)

References PEDERSEN.TBasic Information on Design Features of PIUS.Review of Advanced Light Water Reactor Design Approaches10-13 May, 1994, Moscow, RF, IAEA-TC-879.

PEDERSEN.TA New Generation Nuclear Power Plants.Proceedings ASME/JSME Nuclear Engineering Conference (ICONE-2), SanFrancisco, CA, 1993, March, Vol 2, p 627-631.

Reactorriser

Pressurizersteam volume

Siphon breakerarrangement

Steam

Feed-water

Upperdensity lock

Reactor core

Lowerdensity lock

Figure AIP1PIUS-600: Principle arrangement

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1995-02-13

AIP1:4(4)

Figure A IP IPIUS-600: Reactor building

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1995-02-13

AIP2:i(3)

AIP2

Format

Tide


Reactor type

Power output

Organization (name)

Development

Description

Description

ISER (Inherently Safe and Economical Reactor)

NPP

PWR


JAERI, Japan

Studies

The ISER with a reactor thermal power of 6451

Schedule

PWR type version enclosed in a steel vessel. The use of a steel vessel yieldscertain advantages, since the operating pressure of the primary system can beraised to 16 MPa compared with 9.5 MPa for PIUS-600. The borated waterinventory sufficient for a grace period of 24 hours.

The core utilizes 16x16 array fuel assemblies 2 m in height, and the powerdensity is 61.3 kW/l. Four RCPs pump coolant flows down to the core inletthrough the steam generator. The RCP motors are mounted on the outside ofthe steel vessel head and long pump shafts extend to the impeller on top ofthe SG. In the ISER, the temperature of the inner pool with borated water iskept at about 100 °C.

The ISER has an innovative scram system, similar to the scram density lockvalves system in PIUS. No correspondence to the long-term passive RHRsystem with heat dissipation was proposed.

The normal operation performance and safety responses are rather similar tothose of PIUS, with a few differences due to the integration of the RCS andthe SG into the steel vessel that is submerged in a borated water pool.Obviously, this eliminates some accident situations and affects the decay heatremoval in others.

No ISER plant has been ordered yet. A conceptual design has been prepared,ready for detailed design development.

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1995-02-13

AIP2:2(3)

References

ISER Unit Data

Electrical power

Coolant pressure

Coolant temperatureat the core inletat the core outlet

Boric water temperature

Steam pressure in SG

Average core power deusity

Reactor vessel dimensionheightinner diameter

Fuel assembly array

200 MWe

155 bar

289 °C323 OC

100 °C

57 bar

61.3 kW/l

26.4 m6m

16x16

Safety Aspects of Designs for future Light Water Reactors (Evolutionaryreactors)IAEA, Vienna, Austria, 1993, IAEA-TECDOC-712.

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1995-02-13

AIP2:3<5>

8

10

Figure AIP2ISER: Reactor plant.

I RCP. 2 Pressure relief valve. } Pressuri/cr. 4 Heal exchanger of horatcd water poolcooling system, 5 Steam and feed water pipes. 6 Riser section. 7 Steam generator.8 Pool of boralcd water at I(K) ()C. 9 Core. 10. II Upper and lower density locks.

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1995-02-13

AIP3:i(4)

AIP3

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

ISIS (Inherently Safe Immersed System)

NPP

PWR


Ansaldo Spa, Italy

Studies

The ISIS is a reactor concept of the intejprinciple. Reactor coolant system with the reactor coolant pumps and thesteam generator is housed in a steel reactor pressure vessel to which steamand feedwater pipes are connected. The RPV contains the intermediate poolwith the highly berated water. A reactor tank, provided with wet metallicinsulation, separates the circulating low-boron content primary water coolantfrom the surrounding highly borated water in the intermediate pool. Theprimary loop and the intermediate pool are hydraulically connected at thebottom and top of the tank by means of open-ended tube bundles, the lowerand upper density locks (sec ATP1: PlUS-principle).

The ISIS reactor plant consists of 3 modules of 630 MWt each. The coreconsists of 17 x 17 arrays of reduced length fuel assemblies. The SG is madeof Inconel pipes arranged in two helicoidal bundles, outside the riser annulus.Steam is generated inside the pipes, while primary water flows transversallyon the outside. The two reactor coolant pumps are located at the top of theSG. Motors of these pumps are cooled by highly borated water.

The pressurizer which is installed outside the RPV consists of a hot upperpart with thermal insulation, containing saturated water and steam, and acold lower part containing water at the same conditions as in the inter-mediate pool. The cold and hot parts are hydraulically connected by anumber of vertical pipes to provide mixing in case of surge flows. One pipeconnects directly to the steam dome, acting as a spray in certain situations.The pressurizer and the RPV are hydraulically connected by means of twopipes, one to the bottom and one to the top of the intermediate pool. Thisarrangement improves the natural circulation.

The module is located in a cavity in a strong concrete structure, enclosed in areactor building. The flange of the RPV head is sealed off against an opening

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1995-02-13

AIP3:2(4)

Schedule

in the shaft of this structure, and die volume above it is filled with water,creating a pretty Ian* reactor or plant pool, that will serve as an emergencyheat sink. The upper, hot part of the pressurizer is also located in this pool.Two passive pool cooling loops are connected to air coolers.

The ISIS concept features deal with the absence of:

o

o

o

o

The control rods and their driving system

The safety grade diesel generator set

The safety grade active decay heat removal systems

The safety injection sysuns

The ISIS NPP power plant application can be the following: Electricitygeneration and district heating or water desalination.

No ISIS plant has been ordered yet Significant amounts of design work andverification testing remain to be done.

ISIS Unit Data

Electrical output

Core and reactivity control

205 MWe

Fuel materialFuel inventoryAverage core power densityAverage discharge bumupInitial enrichmentReload enrichment at equilibriumRefuelling frequencyActive core height/diameterNumber of fuel assembliesNumber of fuel rods per assemblyClad material

Reactor coolant systemCoolant pressureCore temperature inlet/outlet

Reactor pressure vesselOverall length/inside diameterVessel material (lining)Gross weight

UO224.3 t26.7 kW/kg U38MWd/kg2.0 - 3.0 %3.5%18 months2.92 m/2.0 m69264Zircaloy4

140 bar271°C/310°C

26.5 m/4.9 mCarbon steel (SS)7821

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1995-02-13

AIP3:3(4)

Steam generatorTypeConfigurationPeed water pressure/temperatureSteam pressure/temperature

Helical tubeVertical52.6 bar/120 °C46 bar/290 °C

References ISIS Reactor System Description and Devdopment status. Integral DesignConcepts on Advanced Water-cooled Reactors.Obninsk, Russia, 16-20 May, 1994, IAEA-622-I3-TC-633.17.

QnettiURizzoFLThe Inherently Safe Imnr rsed System (ISIS) reactorNuclear Engineering and Design, 143.1993. p 295-300.

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1995-02-13

AIP3:4(4)

REACTOR VESSEL /

Figure AIP3ISIS: Reactor module.

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1995-02-13

AIP4:l(3)

AIF4

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

MAP (Minimum Attention Plant)

NPP

PWR


Combustion Engineering, US

Basic design

The MAP design is a PWR with icircuit, which is rated at 300 MWe (the 900 MWe concept was alsodeveloped). The safe MAP operation is ensured by a nearly triple reductionof the average core power density compared with that of the existing PWRs.The heat is removed from the core to the once-through SG by natural circu-lation. The fuel assembly consists of four bundles of standard PWR fuel rods(arranged within a square 9 x 9 array) divided by a water channel, whereabsorbing elements can be inserted. The fresh fuel assembly contains 32burnable poison rods made of Gd2O3 in natural UO2. These rods are perma-nently installed in the fuel assembly. After the first irradiation cycle sixteenburnable poisons of Gd2O3 in the low-neutron-absorbing diluent (AI2O3 orZ1-O2) are inserted into the fuel assembly, which also stay in the fuel assemblyuntil the end of the fuel residence time. The reactivity margin ensured byburnable poisons is 0.08 AK/K.

The reactivity effects resulting from power variations and the fuel burnup arecompensated for by changes in the coolant density. For this purpose, theoutlet coolant temperature is kept equal to the saturation temperature underall operating conditions. The coolant density is controlled by changing thesaturation pressure in the primary circuit, with the upper plenum of thereactor vessel serving as a pressurizer.

The MAP design envisages that the following events would occur withoutoperator action or the reactor shutdown: absorbing rod drop, a feed waterpump trip, switch-off of a steam generator module, switch-off of thecondenser.

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1995-02-13

AIP4:2(3)

Schedule

The residual heat is removed through the secondary circuit The reactorvessel size is such that at the chosen nominal thennal power of 300 MWe, inthe event of the small leak accident accompanied by a loss of heat sink andthe emergency water supply, the core uncovery begins in 2.6 - 6 hoursdepending on the initial coolant volume.

The major departures of the MAP from the traditional PWR design are:

o Incorporation of all primary system components, including thesteam generators, within a single pressure vessel

o Small vessel penetrations which are limited to the feedwaterinlet, steam outlet and control element drive mechanisms (onthe top of the vessel)

o Intrinsic reactor control to compensate for both the long-termeffects of fuel depletion and short-term power manoeuvringrequirements, without any requirement for soluble boron orregulation control rods. The strongly negative moderatordensity coefficient of reactivity allows for power self-regulation; burnable absorber rods provide for long-termreactivity control

o Control rods insertion is required for few hypothetical accidents

No MAP plant has been ordered yet No current development work.

MAP Unit Data

Thermal power

RPV diameter

RPV height

RPV wall thickness

Fuel assembly array

Fuel material

Maximum fuel heating rating

900MWth

5.5 m

24.7 m

292 mm

Square 9 x 9

UO2

170W/cm

References Safety aspects of designs for future light water reactors (evolutionaryreactors).IAEA-TECDOC-712,1993, Vienna, Austria.

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1995-02-13

AIP4:3(3)

feed waterentrance

steam exit

feed waterentrance

steam exit

Figure AIP4MAP: Reactor block diagram.

1 Steam generator module, 2 Core, 3 Pressurizer, 4 Water level in the primary circuit,5 Line of pressure relief and gas blow, 6 CSS drive guides

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1995-02-13

AIP5:?(3)

AIP5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

VPBER-600

NPP

PWR


OKBM (Experimental machinery design bureau), RF

Basic design

Description The VPBER-600 reactor plant with a net electric output 600 MWe is areactor with an integral arrangement of the primary circuit Steamgenerators, pressurizer, main circulation pumps, heat exchangers for theECCS are located in one vessel of about 6 m diameter. An additional"passive" barrier is provided, in the form of the stainless steel guard vessel,enclosing the reactor and primary circuit systems. The reactor plantdevelopment is based on the creation and successful operation of ice-breakerplants as well as AST-500 reactor plant experience.

The VPBER-600 arrangement has the following advantages:

o

o

o

o

o

o

o

Lower probability of primary circuit depressurization

No "large leak" accidents

Elimination of rapid dryout of the core

Reduced effect of the failed steam generator on the primarycircuit

Decrease of neutron fluence

Reactor vessel is not subjected to cold water flooding

Hydroaccumulators, coolant recirculation could be excluded

The core heat removal is achieved with forced coolant circulation by sixcanned circulating pumps which are installed in the reactor vessel bottom.

In the circular gap between the reactor vessel and in-vessel well at the levelabove the core, the steam generator is arranged. The steam generator is aonce-through device consisting of 12 independent sections, each consistingof 18 subsections.

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1995-02-13

AIP5:2(3)

Schedule

Cooling the reactor down in the emergency is achieved by built-in heatexchanger-condensers through an intermediate circuit and by the evaporationof water from two tanks with natural coolant circulation in all heat transferloops for not less than 3 days. The reactor depressurization system isintended in the event of the RPV integrity loss.

No VPBER-600 plant has been ordered yet There is significant design workto be done, before commercial application. No current development work isbeing carried out Today the design team is looking at a large power level forloop type APWR.

VPBER-600 Unit Data

Reactor thermal power

Primary circuit circulation

Primary circuit coolant parametersPressureCore inlet temperatureCore outlet temperature

Core power density

Secondary circuit parameters

1800MWt

Forced

157 bar294.5 °C325 °C

69kW/l

Steam outputSuperheated steam pressureSuperheated steam temperature

Power variation range

Lifetime

Maximum design earthquake, on MSK-64 scale

3 350t/h65 bar305 °C

30-100 % Nnom

60 years

8

References MTTENKOV, F MVPBER-600 enhanced safety water cooled mediumpower reactor.Proceedings International Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 25-29,1992, Tokyo, Japan, v 1, p 4.6.

MJTENKOV, F MVPBER-600 Conceptual features on safety analysis results.Nuclear Safety, 1993, Vol 39, No. 2, p 237-242.

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1995-02-13

AIP5:3(3!

VPBER-600 reactor plant schematic diagram

<i .. Into deaerator"• From condensation

pump

From watertreatment system

Figure A1P5VPBER-600: Reactor plant flow diagram.

1 Main circulating pump2 Reactor3 Steam generator4 Heat exchanger - condenser5 Continuous heat removal system6 Self-actuating devic-s (direct action)7 Intermediate heat exchanger8 CRDM

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910111213141516

Guard vesselContainmentHeat exchangers unitEmergency boron injection systemTank with boron solutionPassive heat removal systemCoolant clean-up and boron reactivity control systemPrimary circuit makeup system


1995-02-13

AIP6:i(4)

AIP6

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SIR (Safe Integral Reactor)

NPP

PWR


ABB-CE, USA; Rolls Royce, UK

Basic design

In the SIR power plant with a theisystem components (including the steam generator, pressurizer and maincirculation pump) are enclosed in a single steel pressure vessel. The safetysystems make extensive use of stored water energy and natural circulation.The size and modular design provide the opportunity for utilities to makeincremental power additions with a low capital investment and a shortconstruction period. A 1 200 MWe station could be provided by having fourreactor modules feeding into a common steam supply system.

The reactor size was limited by the current practice of RPV diameters whichcould be manufactured. This resulted in a power output of 320 MWekeeping the conservatism of the design margins. There is a prospect ofincreasing the power up to 400 MWe.

The integral reactor design incorporates the core, twelve once-through steamgenerators, six canned rotor pumps at a high level in the vessel. The coreconsists of 65 fuel assemblies of size 28.5 x 28.5 cm. The top part of thevessel forms the pressurizer with its electric heaters. Its specific capacity isfive times larger than that of the PWRs. There is a passive spray system inthe pressurizer which takes water from the riser region and sprays it into thesteam space in the pressurizer. The specific inventory of the reactor coolantis three times as large as that of standard large power PWRs. The contain-ment has a novel form of pressure suppression factory where the water forpressure control is contained in 8 tanks connected to the reactor cavity bylarge diameter pipes.

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1995-02-13

AIP6:2(4)

Schedule The SIR reactor was designed to meet safety criteria of the US and UK andwas submitted to US DOE for consideration for funding in the ALWRprogramme. It was not successful. No current development work is beingcarried out

SIR Unit Data

Net electrical output

Core and reactivity controlFuel inventoryAverage core power densityAverage discharge bumupInitial enrichmentActive core heightCore diameterNumber of fuel assembliesNumber of fuel rods in assemblyRod array in assemblyClad materialNumber of control rods or assembliesControl rod neutron absorber material

Reactor coolant systemOperation cooling pressureCore inlet temperatureCore outlet temperature

Reactor pressure vesselLengthDiameterMaterialLining material

Steam generatorNumber of SGConfigurationType

320 MWe

46.lt54.6 kW/l38MWd/kg3.3 - 4.0 %3.47 m2.59 m6543222x22Zircaloy465B4C

155 bar294 oc318 °C

23.8 m5.8 mCarbon steelStainless steel

12VerticalOnce through

References GIBSON, IThe SIR ReactorIntegral Design Concepts on Advanced Water-Cooled ReactorsObninsk, May 16-20,1994, IAEA-TC-633.17.

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1995-U2-13

AIP6:3(4)

t "DM3 i Nozzle (65)

Safety ReliefValve (3)

CEDM ShaftGuide Tube (65)

PressurizerSeparation Plata

IncoreInstrumentNozzle (6)

Sleam GeneratorUpper Support (12)

Feedwater Line (12)

SafetyInjectionNozzle (2)

Sleam GeneratorFlow Plate

Control ElementGuide Tubes (65)

Excore FluxMonitors (8)

Core SupportBarrel

Flow Skirt

Lower Head Bailie

Cooldovrn SprayNozzle (4)

Pressunzer(Upper Head)

ReactorCoolantPump (6)

Steam Line (12)

Vessel Support (6)

SleamGenerator (12)

ReactorVessel

Sisam GeneratorLower Support '12)

Core

VesselSupport Key (6)

IncoreInstrumentGuide Tube (31)

Figure AIP6SIR: Reactor pressure vessel and internals.

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Cut-away View of Reactor and Building

S E C T I 0 N 1

1 Reactor2 Reactor internals storage pool^ Spent fuel pool4 Spent fuel cask pool5 Safety injection accumulators6 Pressure suppression tanks7 Equipment and personnel hatch8 Steam headers9 Natural CRC cooling tank

10 Crane

S E C T I 0 N 2


1995-02-13

AIP7:1(4)

AIP7

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SPWR (System-integrated PWR)

NPP

PWR


JAERI, Japan

Studies

The SPWR is an integral type reaThe steam generator, die reactor coolant pump (RCP), and pressurizer, areincorporated in a reactor pressure vessel (RPV)- The tank with boratedwater is located outside the RPV. This yielded more space inside the RPVand enabled an SG capacity increase and a simplified internals arrangementThe RPV is covered with a water-tight shell with mir-or insulation on itsinside. The RPV with the water-tight shell is installed in a steel containmentvessel filled with borated water. The SPWR uses some technical solutions ofthe PIUS plant design, but the implementation of the SPWR may take amuch shorter time because of its similarity in many respects to existingLWRs.

The core utilizes hexagonal fuel assemblies, some with burnable absorbersmixed with the fuel. Control rods have been eliminated. The canned motor islocated at the top of the riser. The two-loop, helical coil type SG has a once-through design. The pressure head of the RCP is utilised to keep the line tothe poison injection tank closed; if the RCP head is lost, an hydraulicpressure valve will be opened by spring force (borated water enters theprimary loop and shuts the reactor down). An active shutdown system alsouses the hydraulic pressure valve - by closing a valve between the RCP andthe valve. Scheduled shutdowns are performed using traditional active boroninjection.

The emergency injection system is designed to keep the core submerged inwater during all LOCA events. The large water inventory and the eliminationof large breaks reduce the necessary injection capacity. The engineered safetysystems comprise: a two-train Pressure Balanced Injection System (PBIS)which passively injects containment water into the RPV during LOCAevents; an Automatic Depiossurization System (ADS), four trains with eightvalves, which serves to prevent overpressure in the RPV, and to depressurizethe RPV to assist the PBIS during small LOCA events; a four trainContainment Water Cooling System (CWCS) which passively transfers heat

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AIP7:2(4)

Schedule

References

from the containment water to the environment by a heat pipe system and atwo train Residual Heat Removal System (RHRS) which removes decay beatunder normal shutdown and long-term accident conditions.

The SPWR is still in die development by JAERL Experimental and cal-culan'onal studies are being done in support of design activities. No SPWRplant has been ordered yet

SPWR Unit Data

Net electrical output 600 MWe

Core and reactivity controlFuelmaterial UO2Fuel inventory 74.81Average core power density 65.1 kW/lAverage discharge bumup 48 MWd/kgInitial emidunent/reload enrichmentat equilibrium 4.0/4.5 %Moderator material H2OActive core height 2.4 mCore diameter 3.83 mNumber of fuel assemblies 199Number of fuel rods per assembly 325Rod array in assembly TriangleClad material Zircaloy 4

CoolantOperating coolant pressure 138 barCore inlet temperature 288 °CCore outlet temperature 314 °C

Reactor pressure vesselOverall length of assembled vessel 29 mInside vessel diameter 6.6 mGross weight 1 5341

Steam generatorNumber 1Configuration VerticalFeed water pressure 66 barFeed water temperature 210 °CSteam pressure 56 barSteam temperature 295 °C

SAKO.K, OIKAWA, T and ODA, LSPWR (System-integrated PWR)Integral Design Concepts on Advanced Water-cooled Reactors16-20 May, 1994, Obninsk, RF, IAEA-622-I3-TC-633.17

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1995-02-13

AIP7:3(4)

PressurebalancingI i ne

letwelI(N, gas)

surizingI i ne

(Heat pipe type)Pressurizingcap

Containmentwater spray

lineDepres-surizingspray line

PRSS/ PB I S

Poison tank(Highly boratedwater)

Pressurereliefvalve

Water-tishell

Dr i*(fit t:as)

Ccntai vnent vessel

Isolation valve

[NON SAFETY GRADE SYSTEMS]R H R S : Residual heat removal

system (2 trains)H P I S : High pressure

injection system (1 train)

Borated waterfeed Iine

8 0 R H R S /3=3 e v e s

L H P I S

tXX= =0 M S line

0=> FDW l i n e

[SAFETY GRADE SYSTEMS](Passive)

P R S S : Passive reactorshutdown system (2 trains)

P B I S : Pressure balancedinject ion system (2 trains)

C W C S : Containment watercooling system (6 trains)

(Active)Supression pool A D S : Automatic depressur i nation(Borated water) system (4 trams. 8 valves)

Figure AI\J7SPWR:Saf;tysys:ems.

ww viclor/A!J'7 ea

STUDSV1K ECO & SAFETY AB STUDSVIK/ES-y.VKl AIP7:4(4)

FIRM-* AII'7SPWR: Rcat-tor layout.

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1995-02-13

AIP8:i(3)

AIP8

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

B-500 SKDI

NPP

PWR


RRC "Kurchatov Institute", EDO Hydroprcss, RF

Studies

For the B-500 SKDI reactor plant pressure increasover the critical value makes it possible to construct the 515 MWe LWRwith an integral arrangement and with the natural circulation of the coolant inthe vessel with a diameter of less than 5 m. The B-500 SKDI core and steamgenerator are contained within the steel pressure vessel. The pressurizer islocated apart from the pressure vessel. The tube block shroud is 2.8 m indiameter and it separates the riser and downcomer parts of the coolant. Thecore design is based on the contemporary W E R plant technology.

The steam generator is a once-through vertical heat exchanger arranged in anannular space between the vessel and tube block shroud. It consists of 18modules with the titanium alloy tubes of 10.8 m in length, 12 mm in outerdiameter. The total pressurizer volume is 70 nA

In addition to integral arrangement advantages the supercritical coolantpressure results in the following inherent safety features:

o Absence of departure from nucleate boiling on the fuel cladding

o Highly reliable coolant natural circulation

o Affect the reduction of the secondary circuit pressureperturbation on the primary circuit parameters due to a largetemperature difference in the steam generator

o Reduced reactor-specific thermal power due to an efficiencyincrease.

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AIP8:2(3)

Schedule

The core cooling passive safety system consists of the systems which removeheat through steam generators and the secondary circuit system PHRSl andthe primary circuit system PHRS2. They include hydroaccumulators andhydrotanks for core flooding.

No B-500 SKDI plant has been ordered yet No development work is beingcarried out

B-500 SKDI Unit Data

Thermal powerElectric power

Operation pressure at the core outlet

Coolant temperature, °CCore inletCore outlet

Core coolant flow

Time period between refuelling (rated power)

Fuel lifetime

SG steam pressure

Feedwater temperature

Generated steam temperature

Number of steam generator modules

Heat exchange tube material

Tube diameter/thickness

Number of tubes per module

Pitch of SG tube bundle

Calculated effective length of heat exchange tube

Full heat transfer area

1 350 MWt515 MWe

235 bar

365/345 °C*>381.1/378.8 °C*>2470/2880 kg/s*)

2 years

6 years

100 bar

252/240 °C*>

379/375 °C*>

18

Ti alloy

12/1.3 mm

698

21 mm

10.8 m

5 120 m2

(Beginning of fuel lifetime/end of fuel lifetime.)

References/contacts SILIN, V A et alThe integral LWR with natural circulation of the coolant at supercriticalpressure B-500 SKDI.Proceedings International Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 25-29, 1992, Tokyo, Japan, Proceedings, Vol 1.

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1995-02-13

Figure AIP8B-500 SKDI: General layout of the containment arrangements.

1 Reactor2 Pressurizer3 Hydrotank4 Spent fuel pool5 Bubble-condenser

6 Guard vessel7 Hydroacnimulator8 Water stcage tank9 Containnent

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1995-02-13

AIP9:i(3)

AIP9

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

SCLWR

NPP

PWR

Large power/1 100 MWe

University of Tokyo, Japan

Studies

Description The supercritical water is considered to be coolant for direct-cycle SCLWRplant design with a thermal power of 2 703 MW. Compared to the BWR plants,a steam separator, a dryer and recirculation system are not necessary. Thethermal efficiency is high due to the high coolant enthalpy (24 % higher thancurrent BWR plants).

The main features of the SCLWR concept are the following:

o The maximum surface temperature of the stainless steel claddingfor the fuel is conservatively limited to below 450 °C in order toavoid corrosion

o The coolant pressure is 250 bar, well above the critical pressure of

221 bar

o All RPV walls are cooled by the inlet coolant as a PWR vessel

o The control rods are inserted from the top, since there are nosteam separators and dryers

o There are only two coolant loops because of the low mass flowrate

o

o

The diameter of main coolant pipes is smaller than those of PWRs

The main feedwater pumps which are similar to those of a fossil-fuelled plant pump with the coolant above the supercriticalpressure

The turbines are smaller than those of an LWR plant

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AIP9:2(3)

Schedule

o The emergency coie cooling system is similar to that of an ABWR

o The containirrnt is the same as for a BWR

No SCLWR plant has been ordered yet Conceptual design work is under way.

SCLWR Unit Data

Neutron spectrum

Fuel/cladding

Fissile enrichment

Discharge buroup

Thermal/electric power

Thermal efficiency

Inlet/outlet temperature

Coolant flow rate

Pressure

Core height

Average power density

RPV: diameter/thickness

Thermal

UO2/SS

4.2/4.4/4.5 %

42MWd/kg

2703MW/1 100 MW

40.7%

310°C/407°C

2 082 kg/s

250 bar

3.7 m

74kW/l

4.14m/0.34m

References OKA, Y et alSystems design of direct-cycle Super-Critical-Water-Cooled Reactors.Proceedings ENS '94, Vol 2, Paris, 1994, p 473-476.

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STUDSVIK ECO & SAFETY STUDS VIK/ES-95/10 AIP9:3(3)

outlet

coolant density=O.137g/ccZrHi.7 rod fractional % orwater rod fraction=42%

coolant density=O.351g/ccZrHi.7 rod fraction=28% orwater rod fraction=28%

coolant density=0.725g/ccZrHi.7 rod fraction=0% orwater rod fraction=0%

Figure AIP9SCLWR: Reactor pressure vessel.

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1995-02-13

B

BO

Liquid Metal Fast Reactor

Introduction

The development of the fast reactors in the world has a history of more than40 years. Based on this the following conclusions may be drawn:

o The volume of theoretical, technological, designing andexperimental work accomplished in the LMR field is more thanthat accomplished for the substantiation of LWRs and HWRs atthe stage preceding their large-scale commercialization.

o Energy produced by LMR-based NPPs is still more expensivethan that from LWR reactor plants for quite understandablereasons: high enriched fuel, additional intermediate coolantcircuits, additional auxiliary systems (preheating, fireprotection, prevention of water-sodium interaction etc).

o A considerably higher neutron excess for fission in LMRscompared with thermal reactors permits the core design to bechanged using various fuel compositions (metal, oxide, carbide,nitride, cermet) and various structural materials (austenitic andferritic steel) permit LMR to be used, if necessary, for artificialfuel production (Pu, U-233), for incineration of minor actinides(Np, Am, Cm), for the transmutation of harmful long-livedfission products (Tc-99,1-129).

o In the perspective of nuclear power development, the LMRshould not be considered as an alternative to the LWRs butrather as a necessary addition for improving the neutronbalance and efficient use of uranium in the world nuclear powersystem.

o In the decade to come, the LMRs are potential reactors for theeffective utilization of weapon grade plutonium and highradioactive plutonium of LWRs.

During the last years, new radiation-resistant structural materials and fuelcompositions have been developed for LMRs. Some design solutions for theoptimization of reactivity effects and for the incorporation of passive safetymeasures have become known. However, there is still no common opinionon the optimal capacity (large or medium), reactor design, loop or poolarrangement etc.

An LMR base is available today for the design and construction ofevolutionary plants such as large power reactor plants (e.g. Joint EuropeanFast Breeder Reactor and Russian BN-1600), as well as medium power

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1995-02-13

B:2(2)

BN-600M in Russia. A more innovative way is in taking use of simplificationand modulization in the small and medium power range (e.g. AdvancedLiquid Metal Reactor and (ALMR), and Sodium Advanced Fast Reactor(SAFR) in USA; Top Entry Loop Type DEMOFBR in Japan). For bothdirections, the designers use passive safety features mainly through high heatcapacity, thermal conductivity and the natural convection capability of thesodium coolant for decay heat removal.

Also, there are some innovative LMR concepts which use Pb or Pb-Bieutectic as coolant Instead of sodium, the utilization of heavy metal coolantwill allow to increase the fuel breeding ratio (by more than 0.05), to decreasethe void and density reactivity effects, to widen the coolant margins. All ofthese, together with a high natural circulation level and wide possibilities ofreactivity effects means that there is reason to hope for the creation of LMRwith the high inherent safety level.

References

Status of National Programmes on Fast Breeder ReactorsIAEA Publication IWGFR/83,1991.

Carle RemyThe Electricité de France Power Programme.International Conference "Nuclear Energy in 21st century - anenvironmental bonus?". Proceedings, 14/15 April, 1994, Bath,UK.

International meeting on sodium cooled fast reactor safety,Proceedings, 3-7 October, 1994, Obninsk, RF.

Shunsuke KondoDesign trends and major technical issues of advanced non-water nuclear reactors.International symposium on advanced nuclear power systems -design, technology, safety and strategies for their deployment,Proceedings 18-22 October, 1993, Seoul, Republic of Korea.

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1995-02-13

BTSl:i(4)

BTS1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

EFR (European Fast Reactor)

NPP

LMR

Large power/1 500 MWe

EDF, France; Bayernwerlc/PreElectric, UK; ENEL, Italy

Basic design

The EFR project started in 19!UK joined efforts to harmonise their fast reactor development programs. Thedesign companies have rapidly made significant progress in commonunderstanding, common methods and common standards.

The main objectives of the EFR project have been:

o EFRs should be regarded as the lead plant of a commercialseries that could be ordered in the early part of the nextcentury.

o EFR design should retain the basic characteristics of Europeanprototypes (in particular Super Phenix) and develop innovativedesign features for further improvement of safety andeconomics.

o EFRs should achieve parity in economic performance betweenfast reactors and contemporary PWRs.

o EFRs should be licensable in all participating countries with anoverall safety level equivalent to that of future PWRs.

The core has oxide fuel and two configurations are possible in a homo-geneous core 1 m high or a heterogeneous core with a fertile slice just belowthe core mid plane. Both cores have 3 zones with different plutoniumcontents; 207 sub-assemblies in the inner zone and 72 in each of theintermediate and outer zones.

ww victor/BTSl ea

.


1995-02-13

BTS1:2(4)

Schedule

Economic competition is achieved through a compact reactor block,significantly reducing the size and number of structures and components. Asecure design requires a demonstration of robustness against a demandingduty cycle and a high level of confidence in the performance through R&D inthe sodium pool thermo-hydraulics, structural behaviour and materialproperties.

The compactness achieved is demonstrated by the main primary vesseldiameter of 17.2 m (compared to 21 m for Superphenix) largely due to thefollowing:

Simplified internal structures:o Single redano Redan welded on the diagrido Above core structure integrated to the small rotating plugso Compact design of main vessel cooling

Reduced number of main componentso 6 intermediate heat exchangerso 3 primary pumpso External purification

Compact fuel handlingo Intermediate put-down for the central zone of fuel sub-

assemblies using 2 charge machines (direct lift and fixed armmachines)

o Small diameter, solid plate design of rotating plugso In-vessel fuel storage on the diagrid

A number of structural features have also been improved iesponding to thedemands for robustness and improved possibility of inspection, as follows:

o Core support with a second load line within the main vesselimproving fabrication quality, integrity of support and in-service monitoring

o Diagrid main duct nozzle and vessel main cooling pipeconnections are improved with respect to fatigue andratchetting

o Single redan design has been validated for the demanding loadfollowing requirement of the duty cycle

The validation phase of the EFR project is now complete. The next stepshould have been the utilities decision to invest in a FOAK plant but alldecisions have been postponed until the end of 1994.

wwvietor/BTSlea


1995-02-13

BTS1:3(4)

EFR Unit Data

Thermal power

Temperature of coolant at core inlet

Average temperature of coolant at core outlet

Feedwater temperature

Steam temperature

Steam pressure

Primary pump flow rate

Primary circuit pressure drop

Secondary pump flow rate

Secondary circuit pressure drop

Nominal max fuel linear rating

Nominal max fuel clad temperature

Peak bum-up (target value)

Maximum fuel bumup

Fuel life time

3 600MW

395 °C

545 °C

240 °C

490 °C

185 bar

20 172 kg/s

6 bar

15 330 kg/s

4 bar

520W/cm/410W/cm*)

645 °C

20at%

15%

6 years

Stan of life/End of life

References BECCARO, R D, PLESSA, A and HAIGH, PAn Example of International Co-operation:The European Fast Reactor.

MITCHELL, C H and HUBERG, GEFR Programme: Plant Design ActivitiesProceedings International Conference, Oct 2-6,1994, Lyon, France.

The European Fast ReactorNuclear News, 1992, September, p 86-87.

wwvicior/BTSlea

STIDSVIK ECO & SAFETY AB S'lVDSYIK I-IS-s»5 I" BTS1:4,4>

i?ooo-

•'*•": 600C-

Collector

Cr-'d Collector

»Va!l Cooling

v,ei

Denns Trsy

Figure HI SIFiI:K: RciKlor Mruiluivs.

ww vii/lnr/HTSI i.-;i

STUDSVUC i-CO & SAP7"»*. AB STUDSVIK/ES-95/10

1995-02-13

BTS2:i(3)

BTS2

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

BN-600 M

NPP

LMR


OKBM.RF

Detailed design

R&D works on BN-800 ai

Schedule

possibility of commercial power unit creation on the basis of LMFBR, whichwill be competitive in efficiency and safety with advanced PWRs designed atpresent.

The main principles of the advanced BN-600 M concept based on maximumusage of components and systems the reliability of which have beendemonstrated by BN-600 operating experience, use of inherent passive safetysystems, the prevention and mitigation of severe accident consequences bydesign measures. The BN-600 reactor has been operated at BeloyarskayaNPP since 1980, could be considered as a prototype of an advancedcommercial fast reactor. The development showed that the advanced BN-600 M design as compared with the operating BN-600 reactor allows thenext advantages to be achieved:

o Increased safety level

o Improved technical-economical characteristics by 30 - 40 %

o Improved fuel utilization efficiency by 30 - 35 %

o To reduce components and specific weight

o To reduce plant lifetime by up to 50 - 60 years

The possible dates of realization of advanced BN-600 M commercial reactordesign are:The component manufacturing and delivery 1998Commercial NPP with BN-600 M reactor commissioning 2002

ww victor/BTS2 ea

STUDSVnC ECO & SAFETY AB STUDSVDC/ES-95/10

1995-02-13

BTS2:2(3)

References

BN-600 M Unit Data

Reactor capacity

Power output

Primary circuit sodium parametersCore inlet temperatureCore outlet temperature

1440MWt

600 MWe

395 °C550 oc

Secondary circuit sodium parametersSteam generator inlet temperature 530 °CSteam generator outlet temperature 365 °C

Steam parametersPressureTemperature

lifetime

Seismic stability, on MSK scale point

NPP specific weight

137 bar490 °C

50 - 60 years

8

8.9t/MWe

MTTENKOV, F M and KIRUSHIN, A1The Advanced Commercial Fast Reactor BN-600 MProceedings International Conference on Advanced Nuclear Power Plants,October 25-29, 1992, Tokyo, Japan

ww victor/BTS2 ea

.

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 BTS2:3(3)

1995-02-13

secondary loop MCP

steam generatorcontainment

reactor including

reactor core and

primary loop

equipment

Figure BTS2BN-600 M: reactor building.

ww victor/BTS2 ea


1995-02-13

BIS3:i(3)

BIS3

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

DEMO FBR

NPP

LMR


FEPC (Federation of Electric Power Companies), Japan

Basic design

Japan's electricity utility industry will construct a 660 Mtype, demonstration FBR early in the 2000s.

The main targets of these designs are:

o Realization of safety grades comparable with LWRs

o Achievement of a reactor outlet temperature of 550 °C for a

higher plant thermal efficiency

o 1.5 times the construction cost of LWRs

o A High burn-up and long operational cycle aiming at areduction of generating cost

The top-entry loop-type reactor plant is designed so that the primary coolingsystem is composed of the reactor vessel, IHX vessels, and the primarycirculation pump vessels, in which each vessel has a free liquid surface and isconnected with the top-entry reversed U-shape tube pipings to each other,and secondary system is composed of a steam generator and the secondarycirculation pumps.

As for a heat balance, the main steam temperature is set at 495 °C and thepressure at 169 bar based on the evaluation results of reduction in generatingcost, steam generator water-flow stability, etc, at the reactor outlettemperature of 550 °C, resulting in the gross output of approximately670 MWe and a thermal efficiency of 42 %.

A homogeneous 2-region core concept that is adopted in the currentlyoperating or designed reactors, is selected for the initial core stage. Modified

ww victor/BIS3 ea

.


1995-02-13

BIS3:2(3)

Schedule

austenitic stainless steel, which has excellent properties to resist creepstrength and swelling, is used for the cladding tube. 8.5 mm-diameter fuelpins are adopted to guarantee a 15-month operation and a good breedingperformance. Based on burn-up reactivity at the high bum-up core stage, 30control rods are provided to compensate for shift excess reactivity.

The top-entry design of DEMO FBR enables a significant shortening of theprimary piping system and solves the expansion and construction problemsof pipe systems caused by high temperatures. According to estimates thetotal R&D cost for the DEMO FBR unit 1 will consist of almostYen 200 billion.

In January 1994, the top managementof electrical power companies at theirpresidental meeting of FEPC formally decided to launch the construction ofDEMO FBR in 2000s. The construction work on unit 1 will depend on thedevelopment schedule for innovative technologies and operating results ofthe prototype reactor Monju.

DEMO FBR Unit Data

References

Thermal powerNumber of loopsReactor cooling temperatureSteam temperature/pressure

Reactor coreCore typeFuelBum-up

Breeding ratioWith blanketWithout blanket

1600MWt3550 °C495°C/169bar

HomogeneousPu-U mixed oxide90 GWD/t (initial core)150 GWD/t (future highbum-up core)

1.21.05

Utility Industry Decides to Launch Construction of 660 MW ToD-EntrvType DEMO FBR Early in 2000sAtoms in Japan, January, 1994

ww victor/BIS3 ea

_

STUDSYIK HCO& SAFETY AB STl/DSYIK I:S-^5 In BIS3:3<3>

IW5-M2-I3

External vessel chute

Fuel loading/unloading machine

Small rotating plug

Large rotating plug

Root deck

W // /In-vessel chute i

i U i f ' - '

Upper internal structure

Inlet pipingWall thermal protection

linerReactor vessel

Core

Guard vessel

Core support structure

Figure B1S3I)HMC)F;HR: Rc;icti)i stnulinv.

WW vii ldl/HIS 1 i-;i


1995-02-13

BIS4:i(3)

BIS4

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SAFR (Sodium Advanced Fast Reactor)

NPP.WTS

LMR


Rockwell International,Combustion Engineering and others, US

Basic design

The SAFR design is based on the large eexperience in the USA and abroad. It is based on results of supportingdevelopment test activities by ANL's Integral Fast Reactor (IFR)

The SAFR design and economics have evolved on the basis of the lessonslearned from LWR experience. One of the most significant SAFR features ofmaintaining low capital costs is design standardization and certificationlicensing approach.

SAFR is a modular system with a power output of 1 800 MWe. Each450 MWe module consists of one Nuclear Island (reactor assembly andassociated heat transfer equipment) coupled to a single turbine-generatorsystem. The SAFR plant design has a pool arrangement with all itsradioactive primary sodium coolant contained within the reactor vessel.

The reactor system, SAFR, has a heterogeneous core with fuel and blanketassemblies. The mixed uranium-plutonium metal fuel is considered as the fuelload. The SAFR reactor plant has been designed to provide an advancedlevel of safety assurance. For this reason, linear-specific fuel powerdecreased significantly in comparison with traditional values for oxide fuel.The ultimate safety protection mode relies on the inherent responses of theSAFR plant to ensure a safe response to all credible events in addition topostulated accidents without scram. This inherence is made economicallypossible by such distinct SAFR design characteristics as a pool-typeconfiguration and a natural convention decay heat removal systems. Reactordecay heat can be stored for 1 day following reactor shutdown withoutdischarging any heat outside the containment

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1995-02-13

BIS4:2(3)

Schedule

The SAFR reactor design can also be converted for the burning of long-livedminor waste actinides (neptunium, americium and curium) produced inLWRs. As a result, the long-lived nuclides of concern can be efficientlyseparated from spent LWR fuel and destroyed. Under equilibrium conditions,the spent fuel is reprocessed and recycled back to the reactor. For trans-mutation, the minor actinides are also recycled back to the reactor along withthe recycled U and Pu. The reactor operates with a breeding ratio slightlygreater than unity, so that the Pu concentration remains relatively constant.Depleted uranium is added to make up for the heavy metal converted tofission products.

No SAFR reactor plant has been ordered yet No current development workis being carried out

SAFR Unit Data

Thermal powerCore inlet temperature

Core outlet temperature


Core height

Fuel (core/blanket)

Fuel cladding

Number of fuel rods (core/blanket)

Fuel rod diameter (core/blanket)

Maximum linear specific power

Full fuel rod length

Reactor vesselheight (oxide/metal fuel)diameter

Steam outlet temperature

Feed water temperature

1 160 MW350 °C

510°C

3.4 m

1.52 m

UO2-PUO2/UO2; U-Pu-Zr Metal

HT-9

271/127

8.4/12.5 mm

180W/cm

2.2 m

17.9/14.2 m11.7 m

475 °C

257 °C

References MEDIUM-size liquid metal reactor,sodium advanced fast reactor (SAFR),Rockwell international info, 1993.

ww victor/BIS4 ea


1995-02-13

BIS4:3(3)

Simple PoolConfiguration

InherentShutdown

InherentlySafe Core

PassiveDeck Cooling

Passive NaturalConvection Spent

Fuel Cooling

SimpleContainment

Passive NaturalConvection Reactor

Air CoolingSystem (RACS) Very Low

1E PowerRequirements

Passive NaturalConvection

Cavity Cooling

Modified 9Cr-lMoBooster TubeHoneystick SteamGenerator

Common Primaryand SecondaryInducer Pumps

Modified 9Cr-iMoMonoplanarIHTS Piping

Passive SodiumFire Mitigation

Features

Large SeismicMargins

Passive NaturalConvection DirectReactor Auxiliary

Cooling System (DRACS)

Figure BIS4SAFR: Nuclear Island Design Features.

ww viclor/BIS4 ca


1995-02-13

BIS5:i(4)

BIS5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

ALMR (Advanced Liquid Metal Reactor)

NPP.WTS

LMR


General Electric and others, US

Basic design

The Advanced Liquid Metal Reactor (AL(Power Reactor, Innovative Small Module) concept developed by GeneralElectric (GE), while the fuel system is based on the IFR (Integral FastReactor) concept developed by ANL. Major emphasis is placed onsimplicity, inherent safety features and economic operation.

The design approach included the following main elements:

o Compact reactor modules

o Seismically isolated reactor module

o Safety-related equipment limited to nuclear island systems

o Passive shut down heat removal for loss of coolant events

o Passive reactivity control for undercooling and overpowerevents with failure to scram

o Maximum radiation release levels during severe accidents aresufficiently low enough that a formal public evacuation ias notneeded

The reactor module consists of a containment vessel, reactor vessel and itsintemals.

The reactor module, the intermediate heat transport system, and most of thesteam generator system are underground, an approach that has an estimatedcost-benefit in meeting the requirement for radioactivity containment, seismicdesign, sodium fire mitigation, and protection from external threats such assabotage and missiles.

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1995-02-13

BIS5:2(4)

Schedule

The reactor module is about six meters in diameter and 19 meters high. Thecore of about 1 m in height is a radially heterogeneous arrangement of fueland blanket assemblies. The refuelling interval is eighteen months. MetalU-Pii-Zr fuel provided negative reactivity feedback for loss of coolant andtransient overpower events as demonstrated in EBR-II, and the potentiallycompetitive fuel cycle cost with pyroprocessing. The six control rodassemblies are controlled by the Plant Control System.

Primary sodium is circulated in the core by four electromagnetic pumps.The heat is transferred to die intermediate sodium. The single-wall helicalcoil steam generator (HCSG) has been selected for ALMR.

In the unlikely event, when that the intermediate systems become unusableduring power operation, the entirely passive reactor vessel auxiliary coolingsystem (RVACS), which operates continuously, will automatically come intofull operation without the need for operation action. The ALMR conceptsoffer a high level of safety against severe accidents, involving core damage.

The ALMR design can also be converted for the burning of long-lived minoractinides produced in current LWRs.

It must be concluded that the design and licensing activities are far fromcompleted. Conceptual design phase to be completed.

ALMR Unit DataReactors per power block


Turbine throttle conditions (saturated)

Reactor thermal power

Primary sodium temperatureInlet/outlet

Secondary sodium temperatureInlet/outlet

Fuel typeReferenceAlternative

Refuelling interval

Time for breeding

3

465 MWe

66 bar

471 MWt

338 °C/485 °C

282 °C/443 °C

U-Pu-Zr MetalU-Pu Oxide

18 months

50 years, capability

References

ww victor/BIS5 ea

FPPETS.FEeta lDesign of PRISM of Inherently Safe, Economic and Testable Liquid MetalBreeder Reactor Plant.Proceedings International Conference Fast Breeder Systems, USA, 1993,p 13-17.

STimSYIK !•('() & SAFKTY AH HIS5:3(4)

Figure IJIS5Al .MR: KiMiinr iiuuluk1

ww vifliir/HIS5 i:;i

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 BIS5:4i4>

1995-02-13

CONTAINMENTDOME

INTERMEDIATEHEAT EXCHANGER (2)

REACTORVESSEL

REACTORSILO

REACTOR VESSELAUXILIARY COOUNGSYSTEM INLET ANDEXHAUST STACKS

IN-VESSELTRANSFERMACHINE

EM PUMP (4)

CORE-METALFUEL (OXIDEALTERNATIVE)

SeSMIC ISOLATORBEARING

Figure BIS5ALMR: Reactor building.

ww victor/BIS.*) ca


1995-02-13

BIL1

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

BREST 300

NPP

LMR


RDIPE.RF

Studies

With a 300 MWe electricsprogress in the implementation of the innovative safety features specific forthe reactor fuel, lead coolant and other reactor components, as well as someof its design peculiarities.

The BREST 300 reactor core is made up of fuel assemblies of canless designwith high-density nitride fuel. The latter has a high thermal conductivity andagrees favourably with the steel cladding and lead coolant

The canless design of the fuel assemblies and a wide relative spacing of thefuel element lattice provide for the removal of a high level of power bynatural convection (up to 15 % Nnom), as well as the fuel assembly failureby closing one of the flow sections at the core inlet is eliminated due to radialcoolant overflowing in the core.

In order to enhance safety during accidents with a steam-generator pipingrupture, increase of maintainability and the availability factor, reduce vesseldimensions and the volume of the lead circuit - the BREST 300 designconsiders that such a type of reactor layout permits the steam-generators andpumps to be located beyond the boundaries of the main vessel.

With respect to the design, lead circulation is organized in such a v,ay that itsflow rate through the core is determined not by pump pressure bjt thedifference of the level of hot and cold coolant

The system of emergency and operating air cool-down is completely passive,using air tube heat exchangers surrounding the reactor vessel and steamgenerators. Heat is transferred through vessel surfaces and discharged byemission into atmosphere due to natural air thrust.

ww victor/BDLl ea


1995-02-13

BIL1:2(3)

Schedule

Investigations of transients considering extra feedbacks showed that theBREST 300 reactor plant had high self-protection during the followingsevere accidents without the scram:

o Termination of forced circulation in the primary or secondarycircuits

o

o

Self-motion of all the control rods at any power

Coolant cooldown at the core inlet during the secondary circuitrupture

It must be concluded that the design and licensing activities arc far fromcompleted. Conceptual design work is under way.

BREST 300 Unit Data

Thermal power

Net electric power


Core diameter

Core height

Fuel assembly lattice spacing

Core fuel element diameter

Fuel element spacing

Core fuel

Core fa 1 load

LoadofPu/(239Pu+241Pu)

Maximum fuel bum-up

Fuel lifetime

Refuelling intervals

Core breeding ratio

Total breeding ratio

Inlet/outlet lead temperature

Inlet (water)/outlet (steam) temperature

Maximum cladding temperature

Maximum lead velocity

700 MW

300 MW

185(57+72+56)

2.3 m

1.1m

150 mm

9.1;9.6;10.4mm

13.6 mm

UN+PuN

16 t

2.2/1.6t

up to 12 %

5 - 6 years

1-2 years

- 1

- 1

420/540 °C

340/515 °C

650 °C

1.8 m/s

ww victor/B IL1 ea

STUDSVIK ECO & SAFETY AB STUDSVir/ES-95/10

1995-02-13

BIL1:3(3)

References ADAMOV.EOProgress in lead-cooled fast reactor design.Proceedings International Conference on Advanced Nuclear Power Plants,October 25-29,1992, Tokyo, Japan, Vol 2.

+ 11,5

Safeguard

vessel

Steam

generator

013000

Thermal

shielding

Controlrod

Figure KIL1BREST 300: Reactor arrangement.

ww victor/BILl ca


1995-02-13

BIL2:i(3)

BIL2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

LFBR

NPP.TWS

LMR


JAERI, Japan

Studies

The LFBR with an electriiof LMR plant which is being developed by JAERI and others. The designgoals for the LFBR include higher safety, economics and minor actinide(MA) transmutation efficiency than those of MOX-fuel fast reactors.

The core, the core support structure and the primary heat transport systemcomponents are contained in a reactor vessel. The primary heat exchanger isthe steam generator and its helical coil tubes encircle the center part of thereactor which includes the core and the hot coolant plenum. The primarypumps are located at the cold region of the primary coolant flow path. Thecoolant flows along the arrows in the figure.

The nitride thorium fuels are used to bum and transmutate Pu and MA, andto achieve a negative void reactivity. The U-233 of about 1.8 tonnes areproduced and almost the same amount of Pu is transmutated in the bum-uptimes of 3 000 days. MA of 0.8 - 1.4 tonnes generated from 33 - 58 LWRswith 33 MWd/kg are transmutated.

The main safety and design features of the LFBR are the following:

o Reactor coolant paths are designed to prevent Pb-fyO reactionproducts entering into the core in the event of an accident

o The reactor structure can withstand 2G seismic accelerationconditions

o The SG tube is designed as a helical coil type, running along thereactor vessel wall to reduce the reactor vessel diameter

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1995-02-13

BIL2:2(3)

Schedule

o

o

o

o

The lead-cooled SG size is as large as the sodium-cooled one

The pressure drops at the fuel assemblies and the SG aredesigned as low as possible

The decay heat removal system is designed as a water/steamcooled Primary Reactor Auxiliary Coolant System (PRACS)for eliminating lead piping

The electrical prc-heating system is selected

An upper internal structure plug removal is selected for therefuelling system

It must be concluded that the design and licensing activities are far fromcompleted. Conceptual design work is under way.

LFBR Unit Data

Type of plant

Thermal/electrical output

Reactor outlet/inlet temperatureFuelLoaded PuBurn up

Pool type

1 500 MWt/625 MWe

570/470 °C12wt%Pu(U,Pu)1 5N4 tonnes92MWd/kg

Reactor vesselDiameterHeightSystem weight

Thickness

Primary systemHeat exchangerNumber of pumpsLead flow

Intermediate system

Feed water systemFeed water/steam temperatireFeed water flow

9 m17 m9 300 tonnes excluding cover

100 mm

Integrated in the reactor33.70 x 108 kg/h

None

330/530 °C290 x 106kg/h

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1995-02-13

BIL2:3(3)

References TAKANO.HctalA concept of self completed fuel cycle based on lead-cooled nitride-fuel fastreactor.Proceedings ICENES '93, Makuhari, Chiba, Japan, 20-24 Sept, 1993,p 309-315.

ReactorCover Plug

Primary Pump

PRACSCooling Coil

Upper InternalStructure

Core SupperStructure

Control RodDrive Mechanism

(—>o Feed Water

SteamFeed WaterHeaderSteamHeader

LeadCoolant

SteamGenerator

ReactorVessel

Figure BIL2LFBR: Reactor plant arrangement.

ww victor/BIL2 ea

STUDSVUC ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

CO

Gas-Cooled Reactors

Introduction

Graphite-moderated gas-cooled reactors have been operated since 19S6 forcommercial nuclear power. MAGNOX and AGR designs are based onmetallic cladding of the fuel with CO2 as coolant The further developmentof gas-cooled reactors in the USA, Germany, Switzerland, Russia and Japanhas concentrated on the High Temperature Gas Cooled Reactor (HTGR)type using helium as coolant and ceramic cladding of the fuel. The operatingtemperatures of HTGR can be adapted not only to the effective electricitygeneration, but also for industrial cogeneration applications.

The specific safety features of HTGRs are mainly based on:

0 The high temperature resistance of the ceramic core structure

o The large margin between operating temperatures and thefailure limits of the coated particles

o A negative reactivity temperature coefficient and large margin

for allowable temperature rise

o The low power density of the core

0 The large heat capacity of the graphite moderator andstructures

0 The inert helium coolant

The modular HTGR concept has been developed since the 1980s inGermany, the USA, Japan and Russia. These innovative designs incorporatesignificant departures from the large power HTGRs in the reactor coolantsystem arrangement and safety system organization. The design of modularHTGRs is governed by the aim to keep, solely by passive heat transfer, thefuel temperatures so low under all accident conditions that no significantfission product release can take place even including the depressurizationaccident combined with a loss of active cooling. This leads to limitations inpower and power density as well as to a larger surface/volume ratio becausethe decay heat has to be transformed to coolers or structures outside thereactor vessel only by radiation and conduction of the heat via the surface ofthe core and pressure vessel.

The Modular HTR reactor plants are included in the nuclear researchprograms in the USA, Japan, Germany and some other countries.

ww victor/C-gcr ea


1995-02-13

C:2(2)

References

Gas cooled reactor design and safetyVienna, Austria, IAEA, Technical Reports Series N313, 1990.

Shunsukc KendoDesign trends and major technical issues of advanced non-water nuclearreactors.International symposium on advanced nuclear power systems - design,technology, safety and strategies for their deployment. Proceedings18-22 October, 1993, Seoul, Republic of Korea.

ww victor/C-gcr ea


1995-02-13

CI1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

HTR-Modul

NPP, NHPP

HTGR

Small power/80 MWe


Basic design

The HTR-Modul reactor

Schedule

to the time tendency to enhance the unit sizes for further improvements ineconomy as an economic nuclear heat source with high safety margins. Thisconcept is based on the modular NPP approach. One to eight modular unitswith a rating of 200 MWt each can be connected to the one plant (up to640 MWe). The design of modular HTRs is governed by the aim to keep,solely by passive heat transfer (heat conduction, natural convection) the fueltemperatures so low (< 1600 °C) under all accident conditions that nosignificant fission product release can take place even including the depress-urization accident combined with a loss of all active cooling systems.

This leads to limitations in power, power density and radial size of the coreas well as to a larger surface/volume ratio because the decay heat transponedto coolers or structures outside the reactor vessel only by heat radiation andheat conductions. This leads to a relatively large height of the core for asmall diameter, to reduce the temperature gradients in the fuel pebble bed.The reactor is fueled with spherical TRISO fuel elements which pass throughthe core several times until the design burn-up is reached. The possible safetyproblems of the HTR-Modul deal with the accidental ingress of the waterinto the core.

The system and fuel behaviour have been successfully demonstrated also inthe 50 MWt AVR reactor plant Today a HTR-Modul power plant is readyto be constructed and operate commercially.

No HTR-Modul reactor plant has been ordered yet. No current developmentwork is being carried out.

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STUDSVK ECO & SAFETY AB STUDSVnc/ES-95/10

1995-02-13

CI1:2(3)

HTR-Modul Unit Data

Overall PlantNumber of modular unitsReactor thermal powerGeneral output at terminalsNet power output

Reactor Core (one modular unit)Core heightCore diameter, mMean power densityNumber of fuel elementsFuel cycleHeavy metal loading per fuel elementEnrichmentFuel in-core periodFuel burn upCoolant temperature (inlet/outlet)Mean helium pressureHelium mass flow

Steam generator (one modular unit)Live steam pressure/temperatureLive steam mass flow

88x200 MW688 MWe640 MWe

9.6 m3.0 m3.0MW/m3

360 000U/Pu7g7.8%1 000 days80MWd/kg250/700 °C60 bar85kg/s

190 bar/525 °C82kg/s

References Small and medium reactors. II Technical Supplement,Nuclear Energy Agency, OECD, Paris, France, 1991.

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1995-02-13

CI1:3(3)

© Boronolfd

(?) feflftler Rod

© Fut

© Sltoa Otntralor

T) FttdMltr In»

f?) Lrvt SIM* Imt

CiriulalQr

Hat Got Duel

@ Will C*oler

kiw4at<on

Figur CI1HTR-Modul: Nuclear steam supply system.

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1995-02-13

CI2:i(3)

CI2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

MHTGR

NPP, NHPP

HTGR


General Atomics and others, US

Basic design

The MHTGR concept has a unitparison with 200 MWt for HTR-Modul (See CI1). The block type fuel isarranged as an annular cylinder to keep the core temperatures under the1 600 °C limit. The power density can be raised to 6 MW/m3 (for HTR-Modul only 3 MW/m3) due to this geometrical form.

The side by side arrangement is identical to the German HTR-Modul toprevent the ingress of water as well as overheating of metallic componentsand to permit hot stand-by operation with subsequent smooth restarting.Decay heat removal of MHTGR is normally achieved by an additionalcooling circuit.

The MHTGR development strategy is the following:

ooo

No reactor safety systemsNo operator action requiredInsensitive to incorrect operator action

Schedule

The MHTGR can withstand a loss of helium coolant in combination with theloss of all forced circulation from full power without fuel temperaturesreaching the limit level.

The reactor is embedded underground, thus reducing the need for furthersheltering against air plane crash or sabotage. The reactor cavity cooling isdone by natural air convection. In hypothetical accidents the decay heat canalso be absorbed by the surrounding earth.

The design requires a prototype demonstration plant. No MHTGR reactorplant has been ordered yet No current development work.

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CI2:2(3)

MHTGR Unit Data

Configuration (electricity generation)

Thermal power


Thermal power per module

Helium pressure

Helium temperature (cold/hot

Design lifetime

Design basis operation, capacity factor

Other applications

Core height

Core diameter (outer/inner)

Fuel element

Core configuration

Fissile material

Power density

Average enrichment

Fertile material

Control rods

Reserve shutdown channels

Reactor vessel (diameter/height)

Steam generator

Main circulator

Shutdown cooling heat exchanger

Shutdown circulator

4 reactor modules and4 turbine/generator sets

1 800 [4x450] MWt

692 MWe

450 MWt

64 bar

260/690 °C

40 years

84%

Co-generation, process heat

9.45 m

3.2/11.5 m

Prismatic hex-block

84-column annual cylinder

Uranium oxycarbide

5.99 W/cm3

19.9 % U-235

UO2

30 (6 inner, 12 in-core,18 outer reflector rods)

12 (in-core)

6.8/22 m

1, once-through helical,upflow boiling

1, electric motor-driven,magnetic bearings

1, once-through helical

1, electric motor-driven,magnetic bearings

References WILLIAMS, P MMHTGR design status and perspective in the USA.Proceedings Ineternational Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 24-29,1992, Tokyo, Japan, vol 1, p 7.2.1.

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CI2:3L?.

CONTROL R00 ORIVEJRERJEUNG PENETRATIONS

MAINCIRCULATOR

STEAMGENERATORVESSEL

STEAMOUTLET

STEAMGENERATOR

STEEL REACTORVESSEL

ANNULARREACTOR CORE

SHUTOOWNHEATEXCHANGER

SHUTOOWNCOOLINGCIRCULATOR

FEEOWATER J^TINLFI

Figure CI2MHTGR: Reactor Plant Modular.

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1995-02-13

CI3:i(3)

CI3

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

GT-MHR (Gas Turbine-Modular Helium Reactor)

NPP.NHPP

HTGR


Genera] Atomics, US

Studies

The GT-MHR reactor plant with a 600 MW outpuapproach developed by General Atomics. This concept is the result ofcoupling the evolution of a small passively safe modular helium cooledreactor with key technology development in the US during the last decade:large industrial gas turbines, large active magnetic bearings, and compact,highly effective plate-fin heat exchangers.

The main stages of the GT-MHR plant evolution are the following:

o 350 MWtMHTGR Steam Cycle (1985)provides passive safety and standardized modular design

o 450 MWtMHTGR Steam Cycle (1990)retains passive safety and modular designreduces power costs comparable to coal and ALWR

o 550-600 MWt GT-MHR (1994)retains passive safety and modular design (see CI1)

- increases efficiency (48 % vs 32 % f e LWR)reduces thermal and radioactive wastesreduces power costs (as plant output increases)

GT-MHR is showing a promising potential for a significant reduction inpower generating costs. The reactor system is based on the steam cycleMHTGR developed over the last 10 years. It retains many of the key designfeatures including the ceramic coated TRISO fuel, with its capability to retainfission products up to very high temperatures, low power density annularcore, factory fabricated steel vessels, and entirely passive decay heatremoval. The key difference from the steam cycle design is that the GT-MHR uses the Brayton cycle to produce electricity directly with the primaryhelium coolant driving the turbine generator. The precoder and intercooler

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CI3l2(3)

Schedule

are helium-to-water heat exchangers which operate in a very benignenvironment (metal temperatures are less than 121 °C.

No GT-MHR reactor plant has been ordered yet Conceptual design w<~:k isunder way.

GT-MHR Unit Data

Plant parameters

Module power level

Core inlet/outlet temperature

Core inlet pressure

Net efficiency

Reactor parameters

Active core columns

ControlStart-up rodsOperating rodsRSCE channels

Vessel parameters

Inner diameter

Outer diameter at flange

Material

Fuel

Average/maximum temperatureNormal operationAccident

550 - 600 MWt

490 °C/850 °C

70atm

47.7%

102

123628

7.2 m

8.4 m

9Cr-l Mo-V

855 °C/1 200 °C1 270 °C/1 554 °C

References SILADY, F A, NEYLAN, A J and SIMON, W ADesign status of the Gas Turbine Modular Helium Reactor (GT-MHR)Proceedings Nuclear Power & Industry International Conference,27-29 June, 1994, Obninsk, RF

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1995-02-13

CI3:3(3)

Upper Journal/Thrust Bearing

Generator

Recuperator 2 =

HeliumTurbjcompressor

Intercooler

Precooler

Turbine Outlet510°C(950°F)2.64MPa (382psia)

Turbine Inlet850°C(1562°F)7.01MPa(1016psia)

Reactor Inlet490°C(915°F)7.07MPa (1025psia)

Compressor Outlet105°C (222°F)7.24MPa(1049psia)

Compressor Inlet26"C (78°F)2.57MPa (373psia)

Precooler Inlet125°C'257°F)2 59MPa (376psia)

Figure CI3GT-MHR: Power conversion vessel assembly.

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1995-02-13

CI4:i(3)

CI4

Format

Tide


Reactor type

Fower output

Organization (name)

Development status

Description

HTGR-MHD

NPP

HTGR


JAERI, Japan

Studies

Description A conceptual study on a combination of high temperature gas-cooled reactorand magneto-hydrodynamic (MHD) power generators has been considered byJAERI and others aiming at achieving a high efficiency of around 60 % for acore exit gas temperature at 2 300 K. An MHD direct power conversion systemneeds no moving or rotating component exposed to a working fluid at a veryhigh temperature.

The proposed HTGR MHD plant consists of a 1 500 MWt HTGR and fouridentical helium loops. The helium gas serves as both the coolant for the HTGRcore and the working fluid of the MHD power conversion. Each loop has anMHD generator, a regenerative heat exchanger (RHX), a pre-cooler, an inter-cooler, and a turbine-compressor set. The helium gas exits the HTGR at2 300 K, is seeded with cesium, and then enters the MHD generator to produceelectricity. The RHX recovers the heat remained in the generator exit gas todrive the turbine-compressor for gas circulation. An auxiliary cooling system isprovided as an independent means of removing the core decay heat

The HTGR-MGD reactor plant generates electricity with an efficiency of about57%.

The major R&D issues identified for the HTGR-MHD concepts are:

o Further improvement in high-temperature capability and in the life-time of the coated fuel particle (up to - 2600 K)

o Thorough understanding of behaviour of non-equilibrium plasmaand electrode phenomena

Optimization of generator configuration and electrode design

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CI4:2(3)

Schedule

o Development of a helium compressor with a high pressure ratio

and a high-temperature (- 1 500 K) helium turbine

o Development of a large capacity (- 270 MWt), compact RHX

o Development of a large diameter (~ 1.2 m), high-temperature

piping and valves

o Development of a Cs feed and recovery system

Significant design and experimental work remains to be carried out beforeapplication. The conceptual design study is under way.

References

HTGR-MHD Unit Data

Rated reactor power

Fuel type

Pebble diameter

Core diameter/height

Core power density

Core inlet pressure

Core pressure drop

Coolant core velocity

Coolant flow rate

Coolant temperature in/out

Maximum fuel temperature

1 500 MWt

Spherical pebble

50 mm

5.0 m/2.0 m

38.2 MW/m3

7.9 bar

1.5 bar

19.5 m/s

204.8 kg/s

919/2 300 K

2 600 K

TAKIZAKA, T et alAn HTGR-MHD combination for high efficiency power generation.Proceedings ICENES '93, Makuhari Chiba, Japan, 20-24 Sept, 1993,p 297-298.

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CI4:3(3)

HTGRISOOMWt

Inter-Cooler

Compressor

/~\M/G]

Motor-Generator

Cs Feeder

MHD Generator214.8(X4)MWe

RegenerativeHeat Exchanger

Cs Recover

Figure CI4HTGR-MHD: Plant flow diagram.

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1995-02-13

D

DO

Molten Salt Reactors

Introduction

Molten salt reactors (MSR) are fluid fuel reactors. One of the advantages offluid fuel reactors over solid fuel reactors is the ability to carry out on-linefueling and fuel processing. Specifically, the MSRs that are fueled with UF4have the fluoride volatility process available to them, which enables simple,proven, full processing.

Molten salt reactors can be designed with or without on-linr processing, orwith various degrees of processing. For fluid fuel reactors with processing,some of the common nuclear reactor terminology is not applicable. Exceptfor some start-up periods, the reactors are operated at an equilibrium steadystate. The fuel concentration and content do not vary with time. Fissionedfuel is replenished by feeding or by breeding. The term fuel burnup thus hasno meaning as there is no specific amount of energy generation related to aparticular identifiable original amount of fissile material. For the same reason,there is no excess reactivity to compensate for burnup or for the continuouspoisoning caused by fission products accumulation in solid fuel reactors.Also, fluid fuel does not suffer from fuel damage, nor is a meltdown possiblefor a fuel that is already molten. Furthermore, when the fluid fuel itself iscirculated out of the core to an external heat exchanger, fluid fuel reactorscannot have a loss-of-coolant accident, they do not have a separate coolantthat will leave the fuel to generate heat with no cooling. Also decay heatproblems for the MSR arc not as strong in comparison with traditional solidfuel ones because of small, 1- 2 % mole, UF4 concentrations in the fuel salt

MSRs were investigated intensively at Oak Ridge National Laboratory(ORNL, USA) and summarized in a design study of molten salt-breederreactor (MSBR) 25 years ago. Construction material problems and reactorchemical aspects including fission product behaviour have been basicallysolved by using beatifully operated experimental reactor MSRE (ORNL) andmany test loops in ORNL and RRC "Kurchatov Institute" (Russia). Todaythe worldwide research on MSR development is quite small.

This innovative reactor type is under conceptual consideration in the USA,Russia and Japan not only for NPPs, but also for Hgh temperature industrialheat supply and for the utilization of weapon-grade Pu and transmutation ofthe long-lived fission products.

References

MAC PHERSON, HThe Molten Salt Reactor Adventure.Nuclear Sicence Engineering, 1985,90(4), p 374-380.

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D:2(2)

o

FURUKAWA, K and LECOCQ, APreliminary examination on "the next generation" nuclearreactors in comparison with the small thorium molten saltreactor.Tokai University, Japan, 1988.

IGNATIEV, V V et alMolten Salt Reactors: Perspectives and Problems.Energoatomizdat, Moscow, 1990.

GAT, U and DODDS, H LThe source term and waste optimization of molten salt reactorswith processing.Proceedings International Conference Global '93,1993,September 12-17, Seattle, US, Vol 1, p 248-251.

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DI1

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

USR (Ultimate Safe Reactor)

NPP

MSR


ORNL.US

Studies

The USR utilizes an innovatehazards (excess reactivity, decay heat and chemical energy) rather thanpreventing them or mitigating the consequences.

The USR reactor plant with a thermal power of 1 420 MWt is a reactor thateliminates the customary safety concerns of nuclear fission reactors. TheUSR concept has an insignificant source term and no reasonable criticalityaccident. Furthermore, the negligible residual after-heat in the reactorrenders its shutdown capability comparable or superior to conventionalpower sources in that no actions or precautions are required following ashutdown of power.

The USR plant utilizes two principles to achieve ultimate safety. Fissionproducts are continuously removed at the rate that they are produced, thusretaining the inherent source term at an insignificant level. The reactor isoperated with no excess criticality, hence no criticality accident is reasonablypossible. The reactor is safely controlled by its negative temperaturecoefficient. The reactor maintains criticality by an internal breeding ratio thatis trimmed to be exactly one.

To facilitate the continuous fission product removal, the USR reactor plantneeds an extremely short time ( 1 - 6 hours) for processing of idel loading.

The USR plant has a traditional loop arrangement of the primary circuit.Centrifugal RCPs pump LiF-BeF2-ThF4 fuel through the cylindricalhomogeneous core.

The passive inherent safety features of the USR reactor plant make it apotentially economical competitor with high reliability, thereby making it anattractive option for future energy generation.

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DI1:2(3)

Schedule The USR concept is very far from commercial application. Significantresearch and design work need to be done before commercial application. Nocurrent development work is being carried out

USR Unit Data

Fuel salt

Secondary salt

Thermal power

Mass flowrate

Volume flowrate

Inlet temperature

Outlet temperature

Power density

Core volume

Core inner diameter

Core height

Time for salt to passthrough cores

Salt velocity in core

Conversion/Breeding ratio

71.6 LiF-6 BeF2-12 ThF4-0.4 UF4 mol %

8NaF-92NaBF4 mol %

1 420 MWt

5.6 tonne/s

1.7m3/s

510°C

700 °C

100 MW/m3

4.2 m 3

2.6 m

2.7 m

8.3 s

0.33 m/s

1

References GAT, U, and DAUGHERTY, S RThe Ultimate safe (U.S.) reactor.1988, ORNL, USA, DE-AC05-840R2140.

GAT, U, DODDS, H LThe source term and waste optimization of molten salt reactors withprocessing.Proceedings, GLOBAL' 93. September, 12-17,1993, Seattle, Washington,Vol l ,p 248-251.

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DI1:3(3)

PROCESSING PLANTVOLATILE!

NUCLEAR STEAM SUPPLY SYSTEM

-PRIMARYHEAf EXCHANOER

SECONDARY-SALT

STCÄACe. TANKS

Figure 1)11USR: Schematic diagram.

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1995-02-13

DI2:i(3)

DI2

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

FUJI-Pu

NPP.WTS

MSR

Small power/100 MWe

Tokia University, Japan

Studies

The FUJI-Pu reactor pi;

Schedule

plutonium recovered from the spent fuel of LWRs has the followingperformance:

o Fuel self-sustaining characteristics, which means, in practice, aconversion ratio of 0.9. The amount of initial inventory andadditional inventory in the transient stage is also small. It resultsin a few fuel transportation requirements.

o Simple structure, easy operation and maintenance workenhance its safety and economy, because of no opening of thereactor vessel, no fuel assembly fabrication, no fuel-processingin full life except the removal of Kr, Xe and T, achieving aburn-up of 500 %, nearly no control rods required due to thevery low excess reactivity.

For FUJI-Pu the fuel salt composition in the initial load is (in mol %)71.8 LiF-16 BeF2-12 ThF4-0.2 PuF3. The reactor core is of the graphite-moderated type in which the single fuel salt-containing fissile and fertilematerials serve as the core fluid and the blanket fluid.

The temperature of the fuel salt at the outlet of the reactor is 665 °C capableof generating the super heated steam at a pressure of 25 MPa andtemperature of 565 °C. Hastelloy NM is used as the construction material forthe loop type primary circuit

This system could also be effective for TRU-incineration and anti-terrorism.

Significant design work remains to be carried out before application.

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DI2:2(3)

FUJI-Pu Unit Data

Thermal capacityThermal efficiency

Reactor vesselDiameterHeight

Core-zone IMaximum radiusFuel ratio

Core-zone IIMaximum radiusFuel ratio

Blanket zoneThickness (radial direction)Thickness (axial direction)Fuel ratio

ReflectorThicknessFuel ratio

Power densityAveragePeak - initial

Volume of fuel saltIn reactor vesselIn primary system (sum)Flow rate

References M1TACHI. K et al

250 MWt40%

5.95 m4.99 m

127 m7 vol %

1.97 m10 vol %

0.33 m0.40 m20 vol %

0.50 m1 vol%

4kW/lllkW/1

11.2 m3

16.7 m3

0.55 m3/s

Nuti"ar characteristics of a small molten salt power reactor fuelled withplutonium.Proceedings ICENES '93, Makuhar, Chiba, Japan, 20-24 Sept, 1993,p 326-331.

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.

SH DSYIk l( <KV S.\i I TV AH SI'l 1 » \ IK IS

REACT

(FUEL SALT COMPOSITION)

'LrF - BeF, - ThF, - m U F .

- 16 - 12 -0.2<™iv

Kij;urt' 1)12

F l U I - l ' u : K c ; u " t i ) r v n i v s e t ti< in a n d b u i l d i n g .I ( ' l l l l l I ' l l I l u l l s . 2 [ ; l l r l | ) | 11 T i p . i I k - I l l I ' v l l U M l V I . 4 R . M v l l ' l h n

( i ( ' n i l : I ' I I K t u l ' . 7 ( ' i n r . S . l M ) i ; i m p i p e s . I l l R i j j n i M ' S M ' L

i l l l l i - | l l l : - i l i ; i l i - N l l l ' - N . l I U ; i k ' l l ' i

!•! 'I Ir. h M


1995-02-13

DI3:i(3)

DI3

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

MSR-NC

NPP

MSR


RRC "Kurchatov Institute", RF

Studies

The MSR-NC reactor plant witinnovative approach being developed by RRC Kurchatov Institute.

After the MSBR concept was published, new designs of MSR conceptsappeared aimed at reaching a complete self-protection of the reactor. Theyuse the same structural materials as in the MSBR design.

The MSR-NC design differs from MSBR in the complete absence ofcontinuous fuel regeneration, fuel circuit due to the natural convection aswell as to the rearrangement of the core and the circuit. In particular, theMSR permits the graphite as a moderator to be completely excluded andonly to be used as a reflector. In addition, the integrated arrangement of thefuel circuit is proposed, which facilitates solving the problem of radioactivityrelease from the primary circuit.

As applied to the MSR-NC with the homogeneous core, the self-protectionis reached due to:

o Negative effects of reactivity resulting from the increase in thefuel-coolant temperature (practically the only component of thecore), a = 6.42 • 10'5/°C

o Use of the free convection of the LiF-BeF2-ThF4-UF4 fuel inthe nominal power

High heat-accumulating properties of the primary circuit

Use of a passive emergency heat exchanger permitting thedecay heat to be removed from the outer surface of the reactorvessel to the environment

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DI3:2(3)

Schedule No current development work is being carried ouL

MSR-NC Unit Data

Power output

Fuel composition

Coolant salt composition

Fuel salt inlet/outlet temperatures

Coolant salt inlet/outlet temperatures

Reactor vessel outside diameter and height

Core diameter and height

Graphite reflector thickness

Fuel flow rate

Average fuel circuit power density

1 070 MWt

71LiF-27.9BeF2-lThF4-0.075UF4mol %

92NaBF4-8NaFmol%

901 K/l 023 K

699 K/883 K

5.1/18 m

4/4 m

0.4 m

3.55 t/s

7.2W/cm3

References IGNATEV.VVeta lOn the concept of Enhanced Safety Reactors with molten salt fuel andcoolant.Fusion technology, 1991, Dec, v 20, N4 part 2, p 620-627.

IGNATEV, V V et alOn molten salt reactors inherent safety in design and beyond designaccidents.Proceedings ICENES '93, Makuhari, Chiba, Japan, 20-24 Sept 1993,p 332-333.

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DI3:3(3)

Figure 1)13MSR-NC: Reactor plant arrangement.

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Appendix 1:1

List of Abbreviations and Glossary of Terms

A

ABB

ABWR

Accelerator

Actinides

Active component

Active system

ADP

ADS

AEC

AECL

AGR

ALARA

ALMR

ALSEP

ALWR

AMSB

amu

ANL

AP

APWR

ATR

ATWS

Availability Factor

AWTS

Asca Brown Bovcri

Advanced Boiling Water Reactor, see BWR

A device that increases the velocity and energy of charged particles suchas electrons and protons; also referred to as a particle accelerator. In a"linear" accelerator, particles are accelerated in a straight path

The elements with atomic numbers above 88 (actinium, element 89). Theactinides series includes uranium, atomic number 92, and all the man-made transuranic elements

Any component that is not passive is active

Any system that is not passive is active

Advanced Double Pool reactor

Automatic Depressurization System

(U.S) Atomic Energy Commission

Atomic Energy of Canada Limited. A designer/supplier of nuclearreactors

Advanced Gas-Cooled Reactor

As low as reasonable achievable

Advanced Liquid Metal Reactor

Apollo Lunar Scientific Experiment Package

Advanced Light Water Reactor, see LWR

Accelerator Molten Salt Breeder

Atomic mass unit

Argonne National Laboratory, US

Advanced Passive

Advanced Pressurized Water Reactor, see PWR

Advanced Thermal Reactor: A heavy water moderated, light water-cooled reactor

Anticipated Transients Without Scram

The availability factor of a nuclear unit or station is the ratio of timewhen energy can be produced to the total time

Accelerator-driven Waste Transmutation System

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Appendix 1:2

B

Base Load

BDA

BMDO

BN

BNL

BOL

BOM

BOP

Breeder Reactor

Breeding Ratio

BWR

The minimum load produced ever a given period. A station used forbase load is a station that is nrnnally operated when available to providepower to meet the minimum load demands

Beyond Design Accident

Ballistic Missile Defence Organization, US

Russian version of sodium cooled Fast Breeder Reactor

Brookhaven National Laboratory

Beginning Of life

Beginning Of mission

Balance Of Plant

A nuclear reactor that produces more fissile material than it consumes.In fast breeder reactors, high-energy (fast) neutrons produce most of thefissions, while in thermal breeder reactors, fissions are principally causedby low-energy (thermal) neutrons. See Fast rer.ctor

The conversion ratio when it is greater than unity. A high breeding ratioresults in a short doubling time

Boiling-Water Reactor. A light-water reactor that employs a directcycle; the water coolant that passes through the reactor is converted tohigh pressure steam that flows directly through the turbines

CANDU

Capacity Factor

CAREM

CCV

CDA

CE

CEA

CEC

Charge of a reactor

CHP

CANadian Deuterium Uranium reactor: A type of heavy water reactor

See Load Factor

Conjunto Argentina de Reactores Modulares

Cooldown Control Valve

Control Drum Actuator

Combustion Engineering

Commissariat å l'Energie Atomique

Commission of tne European Communities

The fuel placed in a reactor

Combined Heat and Power

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1995-02-13

Appendix 1:3

Conversion as used inreactor technology

Conversion, chemical

Converter reactor

Coolant

COPUOS

Critical

CRU

CWCS

The ratio between the number of fissile nuclei produced by conversion tothe number of fissile nuclei destroyed. If the ratio for a given reactor isgreater than one, it is a breeder reactor, if it is less than one, it is aconverter reactor

The operation of altering the chemical form of a nuclear material to aform suitable for its end use

A reactor that produces some fissile material, but less than it consumes

The medium in a nuclear reactor which removes heat from the reactorcore where fission occurs and heat is produced, and transfers it tosystems which convert the heat into steam

(UN) Committee On Peaceful Uses of Outer Space

Capable of sustaining a nuclear chain reaction

Combined Rotating Unit

Containment Water Cooling System

D

DARPA

DBA

DEC

Decay Heat

DIPS

DNHPP

DOE

Defense Advanced Research Project Agency

Design Basis Accident

Direct Energy Conversion

The heat produced by the decay of radioactive nuclides

Dynamic (Radio)isotope Power System

Decentralized Nuclear Heating Power Plant

US Department of Energy

E

E.MAD

EASEP

ECCS

EFR

EHRS

EM

EMP

Engine Maintenance, Assembly and Disassembly

Early Apollo Scientific Experiment Package

Emergency Core Cooling System

European Fast Reactor

Emergency Heat Removal System

Electromagnetic

Electromagnetic Primary Pump

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1995-02-13

Appendix 1:4

EMP

Enriched fuel

Enrichment

Fast neutrons

Fast reactor

Electromagnetic Pump

Sec Fuel, enriched

i) the fraction of atoms of a specified isotope in a mixture ofisotopes of the same element when this fraction exceeds that inthe naturally occurring mixture;

ii) any process by which the content of a specified isotope in anelement is increased

EOL

EOM

EOT

EPA

EPR

EST-1

eV

EVA

EWST

F

Fail-safe

End Of Life

End Of Mission

Earth Orbital Terminal

(U.S.) Environmental Protection Agency

European Pressurized Reactor

Engine Test Stand Number One

Electron volt

Extravehicular Activity

Emergency Water Storage Tank

The term describes the behaviour of a co

Faul t-/er ror-toleran t

FBR

FDA

FEPS

failure (either internal or external). If a given failure leads directly to asafe condition, the component or system is fail-safe with respect to thatfailure

See Neutron, fast

A nuclear reactor in which no moderator is present in the reactor core orreflector. So the majority of fissions are produced by fast neutrons. If afertile species is present in the fast reactor core or in the blanketsurrounding the cere, it will be converted into fissile material by neutroncapture. When mo:e fissile material is produced than is used to maintainthe fission chain, the reactor is called a breeder

(also called forgiving). The term fault-/error-tolerant, also calledforgiving, describes the degree to which equipment faults/humaninaction (or erroneous action) can be tolerated

Fast Breeder Reactor (see Fast reactor)

Final Design Approval

Federation of Electric Power Companies

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Appendix 1:5

FFTF

Fissile

Fission products

Fissionable

FOAKE

Foolproof

Fossil fuel

FP

FPSE

FSAR

Fuel cycle

Fuel, enriched

Fuel, nuclear

Fuel reprocessing

Fast Flux Test Facility

i) of a nuclide: capable of undergoing fission by interaction withslow neutrons;

ii) of a material: containing one or more fissile nuclidcs

Nuclides produced either by fission or by the subsequent radioactivedecay of the nuclides thus formed

i) of a nuclide: capable of undergoing fission by any process;

ii) of a material: containing one or more fissionable nuclidcs

Rrst-Of-A-Kind-Engineering

Safe against human error or misguided human action

A term applied to coal, oil and natural gas

Fission Product(s)

Free Piston Stirling Engine

Final Safety Analysis Report

The sequence of processing, manufacturing, and transportation stepsinvolved in producing fuel for a nuclear reactor, and in processing fueldischarged from the reactor

Nuclear fuel containing uranium which has been enriched in one or moreof its fissile isotopes or to which chemically different fissile nuclides havebeen added

Material containing fissile nuclides which when placed in a reactorenables a self-sustaining nuclear chain to be achieved

The chemical or metallurgical treatment of spent (used) reactor fuel torecover the unused fissionable material, separating it from radioactivewaste. The fuel elements are chopped up and chemically dissolved.Plutonium and uranium and possibly other fissionable elements are thenseparated for further use

G

GA

GCR

GE

GEO

GHR

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General Atomic

Gas-Cooled Reactor

General Electric

Geostationary Earth Orbit, Geosynchronous Orbit

Gas-Cooled Heating Reactor

STUDSVTX ECO & SAFETY AB STUDSVnC/ES-95/10

1995-02-13

Appendix 1:6

G IS

GLFC

GmbH

Grace period

GSOC

GWe

Graphite Impact Shell

Graphite Lunar Module Fuel Cask

German term signifying "Limited"

The period of time which a safety function is ensured without thenecessity of personnel action in the event of an incident/accident

Geosynchronous Orbit Complex

Gigawatt (109 watts) electric

H

Half-life

Heavy water

HEO

HEU

HR

HSBWR

HTR or HTGR

HTTR

HWR

HX

Hydropress EDO

The period of time required for the radioactivity of a substance to dropto half its original value; the time that it takes for half of the atoms of aradioactive substance to decay. Measured half-lives vary from millionthsof a second to billions of years

Deuterium oxide (D2O): water containing significantly more than thenatural proportion (1 in 6500) of heavy hydrogen (deuterium) atoms toordinary hydrogen atoms

High Earth Orbit

Highly-Enriched Uranium

Heat Rejection

Hitachi Small Boiling Water Reactor

High-Temperature Gas-Cooled Reactor. A graphite-moderated, helium-cooled advanced reactor that utilizes low enriched uranium

High Temperature Test Reactor

Heavy-Water Reactor. Heavy water is used as a moderator in certainreactors because it slows down neutrons effectively and also has a lowcross section for absorption of neutrons

Heat Exchange

Reactor Plant Design Bureau, Podolsk, Russia

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1995-02-13

Appendix 1:7

I&C

IAEA

ICRP

IEA

IFR

rasHASA

INET

Inherent safetycharacteristics

INSRP

Isotopes

IUS

Instrumentation & Control system

International Atomic Energy Agency

International Commission on Radiological Protection

International Energy Agency

Integral Fast Reactor

Intermediate Heat Exchange

International Institute for Applied System Analysis

Institute Nuclear Energy Technology, China

Safety achieved by the elimination of a specified hazard by means of thechoice of material and design concept

Interagency Nuclear Safety Review Panel

Nuclides having the same atomic number (i.e. identical chemicalelement) but different mass numbers; e.g. 92-Uranium-235 and92-Uranium-238. Isotopes have the same chemical properties butslightly different physical properties

Inertial Upper Stage

JJAERI Japan Atomic Energy Research Institute

K

KEPCO

keV

kT

kW

kWh

Kansai Electric Power Company

Thousands electron volts

Kiloton

Kilowatt

Kilowatt hour (103 watt hour)

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1995-02-13

Appendix 1:8

LANL

LCP

LEO

LET

LEU

LH2

LiH

LINAC

Linear accelerator

LLFP

LMCPBR

LMFBR

Load Factor

LOCA

LOF

LOHS

LOT

LRC

LWGR

LWR

Los Alamos National Laboratory

Large Communications Platform

Low Earth Orbit

Linear Energy Transfer

Low Enrichea Uranium

Liquid hydrogen

Lithium hydrogen

Linear Accelerator

A long straight tube (or series of tubes) in which charged particles(ordinarily electrons or protons) gain in energy by action of oscillatingelectromagnetic fields

Long Lived Fission Products

Liquid Metal Cooled Pebble Bed Reactor

Liquid Metal Fast Breeder Reactor. A fast reactor that employs liquidmetal (sodium) as a coolant The sodium in the primary loop passesthrough the reactor and transfers its heat to sodium in a secondary loop.This sodium then heats water in a tertiary loop which produces steamand drives a turbine. See also Fast Reactor

The load factor of a nuclear unit or station for a given period of time isthe ratio of the energy that is produced during the period considered tothe energy that it could have produced at maximum capacity undercontinuous operation during the whole of that period. Also calledCapacity Factor

Loss Of Coolant Accident

Loss Of Flow

Loss Of Heat Sink

Lunar Orbital Terminal

Lunar Resources Complex

Light Water Graphite Reactor. A reactor that uses ordinary water as acoolant and graphite as a modertor and utilizes slightly enrichedUranium-235 (e.g. Russian multi-channel RBMK reactor plant)

Light-Water Reactor. A nuclear reactor that uses ordinary water as botha moderator and a coolant and utilizes slightly enriched Uranium-235fuel. There are two commercial LWR types: the boiling-water reactor(BWR) and the pressurized water-reactor (PWR)

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STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 Appendix 1:9

M

1995-02-13

MAP

MARS

MCP

MeV

MHTGR

MHW

Minor actinides

Moderator

MOX fuel element

MPD

MSBR

MSPWR

MSR

MW

MWe

MWh

MWt

Minimum Attention Plant

Multipurpose Advanced Reactor Inherently Safe

Main Circulation Pump

Million electron volts

Modular HTGR, see HTR

Multihundred watt

The transuranic clemcr iiinus plutorium. Usually this term is used torefer to neptunium. .. • . ncium, and curium. Some also refer to these asthe "minor" tnur . ,\cs. Plutonium is the dominant transuranic, butthese minor tra .anics contribute comparable radioactivity in spentfuel

A material, such as ordinary water, heavy water, beryllium, graphite andsome others used in a nuclear reactor to slow down fast neutrons sofissile nuclei can more easily and efficiently capture them, thus increasingthe likelihood of further fission

Mixed Oxide fuel element Fuel element in which fuel is an intimatemixture of uranium and plutonium oxides

Maximum Permissible Dose

Molten Salt Breeder Reactor, s% MSR

Mitsubishi Simplified PWR

Molten Salt Reactor. A nuclear reactor that uses fluk fuel, e.g. moltensalt fluorides of Li, Be, Th, U.

Megawatt

Megawatt (1CK> watts) electric

Megawatt hour (10^ watts hour)

Megawatt (106 watts) thermal

N

NaK

NASA

NC

NCV

Sodium-potassium eutectic mixture

(U.S.) National Aeronautics and Space Administration

Natural Circulation

Nozzle Control Valve

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1995-02-13

Appendix 1:10

NEA

NEP

NEPSTP

NERVA

Neutrons, fast

Neutrons, slow

Neutrons, thermal

NF

NHP

NHPD

NPP

NPPS

NRC

NRDS

NSSS

NTS

Nuclear energy

Nuclear fuel

Nuclear reaction

Nuclear reactor

Nuclide

OECD Nuclear Energy Agency

Nuclear Electric Propulsion

Nuclear Electric Propulsion Space Test Program

Nuclear Engine for Rocket Vehicle Application

Neutrons of kinetic energy greater than some specified value. This valuemay vary over a wide range and will be dependent upon the application,such as reactor physics, shielding, or dosimetry. In reactor physics thevalue is frequently chosen to be 100 000 eV (electron-Volt)

Neutrons of kinetic energy less than some specified value (see neutrons,fast). In reactor physics the value is frequently chosen to be 1 eV

Neutrons in thermal equilibrium with the medium in which they exist

Nuclear Furnace

Nuclear Heating Plant

Nuclear Heat and Power Plant

Nuclear Power Plant A reactor or reactors together with all structures,systems and components necessary for the production of power (i.e. heator electricity)

Nuclear Power Propulsion System

(Nuclear Regulatory Commission): US body regulating the use ofnuclear energy

Nuclear Rocket Development Station

Nuclear Steam Supply System

Nevada Test Site

Energy released in nuclear reactions or transitions

See Fuel, nuclear

A reaction involving a change in an atomic nucleus, such as fission,fusion, neutron capture, or radioactive decay, as distinct from a chemicalreaction, which is limited to changes in the electron structuresurrounding the nucleus

A device in which a fission chain reaction can be initiated, maintained,and controlled. Its essential component is a core containing fissionablefuel. It is sometimes called an atomic "furnace"; it is the basic machine ofnuclear energy

Any species of atom that exists for a measurable length of time. Theterm is used synonymously with isotope. A radionuclide is a radioactivenuclide.

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STUDSVIK ECO & SAFETY AB STUDSVDC/ES-95/10

1995-02-13

Appendix 1:11

O

OECD

OKrtM

OMEGA

OTV

Organisation for Economic Co-Operation and Development

Reactor Plant Design Bureau, Nizni-Novgorod, Russia

Options for Making Extra Gains from Actinidcs and Fission Products

Orbital Transfer Vehicle

Passive component

Passive system

PBIS

PbTe

PC

PCCS

PCI

PCIV

PCRV

PCS

PCV

Peak load

Pebble bed reactor

Per capita

PHRS

PIUS

Plutonium

PPO

A component which does not need any external input to operate

Either a system which is composed entirely of passive components andstructures or a system which uses active components in a very limitedway to initiate subsequent passive operation

Pressure Balanced Injection System

Lead telluride

Power Convenor

Passive Containment Cooling System

Pellet Clad Interaction

Prestressed Cast Iron Vessel

Prestrcssed Concrete Reactor Vessel

Power Conversion System

Pressure Containment Vessel

The maximum load produced by a unit or group of units in a statedperiod of lime. A station used for peak load generation is a station that isnormally operated to provide power during maximum load periods only

A reactor, which utilizes spherical fuel elements

Per unit of a population

Passive Heat Removal System

Process Inherent Ultimate Safety Reactor

A heavy, radioactive, man-made metallic element with the atomicnumber 94, created by absorption of neutrons in uranium-238. Its mostimportant isotope is plutonium-239, which is fissile

Pressed Plutonium (-238) Oxide

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STUDS VIK ECO & SAFETY AB STUDSVncyES-95/10

1995-02-13

Appendix 1:12

Primary energy

Proton

PSA

PSI

PSM

PSOV

PUC

PUREX process

Pyroprocessing

PWR

The energy content of fuels before they are processed and converted intoforms used by the consumer. Primary energy refers to energy in the formof natural resources: water flowing over a dam, freshly mined coal,crude oil, natural gas, natural uranium. Only rarely can primary energybe used to supply final energy; one of the few forms of primary energythat can be used as final energy is natural gas

A particle with a single positive unit of electrical charge and a mass thatis approximately 1.840 times that of the electron. It is the nucleus of thehydrogen atom and a constituent of all atomic nuclei.

Probabilistic Safety Analysis

Paul Scherrer Institute, Switzerland

Power System Module

Propellant Shutoff Valve

Public Utility Commission

The plutonium and uranium extraction (PUREX) process is an aqueousprocess used in several foreign commercial and U.S. defense programsfor separating out elements in spent nuclear fuel

Nonaqueous processing carried out at high temperatures. An example ofthis is the relatively new technology being developed for reprocessing

Pressurized-Water Reactor. A light-water moderator and cooled reactorthat employs an indirect cycle; the cooling water that passes through thereactor is kept under high pressure to keep it from boiling, but it heatswater in a secondary loop that produces steam that drives the turbine

R

R-MAD

rad

Radioactive

Radioactive decay

Reactor Maintenance, Assembly and Disassembly

Radiation absorbed dose

Referring to the spontaneous transformation of one atomic nucleus intoa different nucleus or into different energy states of the same nucleus

The spontaneous transformation of one atom into a different atom orinto a different energy state of the same atom. The process results in adecrease, with time, of the original number of radioactive atoms in asample

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1995-02-13

Appendix 1:13

Radioactive waste

Radioisotope

RBE

RBMK

RBR

RCC

RCP

RCS

RDIPE

Reactor core

Recycling

rem

Reprocessing, fuel

RHR

RJPV

RRA

RRC-KI

RTG

RVACS

The unwanted radioactive materials formed by fission and other nuclearprocesses in a reactor or obtained in the processing or handling ofradioactive materials. Most nuclear waste is initially in the form of spentfuel. If this material is reprocessed, new categories or waste result: high-level, transuranic, and low-level wastes (as well as others)

A radioactive isotope. An unstable isotope of an element that decaysspontaneously, emitting radiation. Radioisotopes contained in the spentfuel resulting from the production of nuclear power generally fall intotwo categories: fission products and transuranic elements (known astransuranics, actinides, or TRU). and activation products produced byneutron absorption in structural material in the spent fuel

Relative Biological Effectiveness

See LWGR

Rotating Bed Reactor

Reinforced Concrete Containment

Reactor Coolant Pump

Reactor Coolant System

Research and Development Institute of Power Engineering, Moscow,Russia

The central portion of a nuclear reactor containing the fuel elements andthe control rods (and part of the coolant and moderator), where most ofthe energy is produced

The reuse of fissionable material, after it has been recovered by chemicalprocessing from spent reactor fuel

Roentgen equivalent man

A generic term for the chemical and mechanical processes applied to fuelelements discharged from a nuclear reactor, the purpose is to removefission products and recover fissile (uranium-233, uranium-235,plutonium-239), fertile (thorium-232, uranium-238) and other valuablematerial

Residual Heat Removal

Reactor Pressure Vessel

Rolls-Royce and Associates

Russian Research Centre "Kurchatov Institute", Moscow, Russia

Radioisotope Thermoelectric Generator

Reactor Vessel Auxiliary Cooling System

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1995-02-13

Appendix 1:14

SADS

Safeguards

SAFR

SBWR

SDR

SECURE

SES

SCLWR

SG

SiGi

Simplified design

SIR

SIS

SMF

SMR

SMS

SNAP

SNPS

SNR

SPAR

Spent fuel

SPR

SPS

STA

Subcritical Accelerator-Driven System

Tenn used to refer to the total set of international verifications,observations, etc, which together constitute a determination that nuclearmaterials (or, in some international agreements, facilities or othermaterials) have not been diverted from nuclear power programmes tothe production of nuclear weapons

Sodium Advanced Fast Reactor

Simplified or Small BWR. See BWR

Slowpoke Demonstration Reactor, see SES

Safe Environmentally Clean Urban Reactor

Slowpoke Energy System

Steam Cooled Light Water Reactor

Steam Generator

Silicon germanium

A system designed with a minimum number of components to achievethe related safety function and relying as little as possible on support

sy.rtf-ms

S i , ntegral Reactor

jv / Injection System1 n t Manufacturing Facility1 cc ill and Medium-sized Reactors

•>' jp Mobile System

.bpace Nuclear Auxiliary Power program, US

"Space Nuclear Power Source

Space Nuclear Reactor

Space Power Advanced Reactor

Nuclear reactor fuel that has been irradiated (used) to the extent that it:an no longer effectively sustain a chain reaction and therefore has beenremoved from the reactor for disposal. This radiation fuel containsfission products, uranium, and transuranic isotopes

Advanced Space Nuclear Power Program

Satellite Power System

Science and Technology Agency, Japan

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STUDSVIK ECO & SAFETY AB STUDSVIKyES-95/10

1995-02-13

Appendix 1:15

STAR-C

STS

Subcritical

Space Theimionic Advanced Reactor Compact, US

Space Transportation System (Space Shuttle)

Not capable of sustaining a nuclear chain reaction, but involving somedegree of multiplication of neutrons

T-l l l

TAGS

Target

TBCV

TCT

TE

TFE

TI

TITR

TOP

TOSBWR

TPC

TPS

Transmutation

Transuranic

TRIGA

TRU

TRUEX

TSCV

Tantalum alloy (Ta-8W-2Hf)

Silver antimony germanium telluride

Material subjected to panicle bombardment (as in an accelerator) inorder to induce a nuclear reaction

Turbine Bypass Control Valve

Thermionic Critical Technology program, US, see SNP

Thermoelectric

Thermionic Fuel Element

Thermionic

Thermionic Test Reactor

Transient Over Power

Toshiba Simplified Boiling Water Reactor

Transmutation Plant System

TRIGA Power Safety System, see DNHPP

The transformation (change) of one element into another by a nuclearreaction or series of reactions

An element above uranium in the Periodic Table of elements - that is,one that has an atomic number greater than 92. All transuranics areproduced artificially (during a man-made nuclear reaction) and areradioactive. They are neptunium, plutonium, americium, curium,berkelium, californium, einsteinium, fermium, mendelevium, nobelium,and lawrencium

Training of personnel, nuclear Research, Isotope production, GeneralAtomic

Transuranium elements

A chemical solvent process under development to extract transuranicsfrom high-level waste

Turbine Shutoff Control Valve

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STUDSVIK ECO & SAFETY AB STUDSVKyES-95/10 Appendix 1:16

i 995-02-i 3

TVA

TWh

TWS

TZM

Turbine Valve Actuator

Terrawatt (1012 watt) hour

Transmutation Waste System

Molybdenum alloy (Mo-0.6, Ti-0.1, Zr-0.035)

uU-ZrH

UO2

Uranium

USR

Uranium zirconium hydride

Uranium dioxide

A radioactive element of atomic number 92. Naturally occurring uraniumis a mixture of 99.28 per cent uranium-238,0.715 per cent uranium-235,and 0.0058 per cent uranium-234. Uranium-235 is a fissile material andis the primary fuel of nuclear reactors. When bombarded with slow orfast neutrons, it will undergo fission. Uranium-238 is a fertile materialthat is transmuted to plutonium-239 upon the absorption of a neutron.

Ultimate Safety Reactor, see MSR

V

W E R Russian version of PWR plant

wwWANO

Waste separation

WEC

WOCA

WRI

Westinghouse

World Association of Nuclear Operations

The dividing of waste into constituents by type (for example, high-level,medium-level, low-level) and/or by isotope (for example, separating outplutonium and uranium). The waste may be separated by a chemicalsolvent process such as PURES or by any of a number of other chemicalor physical processes

World Energy Council

World Outside CPE Areas (MEM ou Monde å Economic de Marché)

World Resource Institute

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STUDSVIK ECO & SAFETY AB STUDSVKyES-95/10

1995-02-13

Appendix 1:17

X

XE Experimental Engine

zZrH Zirconium hydride

References

Small and medium reactorsI. Status and prospectsReport OECD, NEA, 1991, Paris, France.

Safety related terms for advanced nuclear plantsReport IAEA-TECDOC-626,1991, September.

Developing Technology to Reduce Radioactive Waste May Take Decades and BeCostlyReport G AO/RCED-94-16,1993, December.

Space nuclear power systems 1989Orbit book company, Inc, 1992, Malabar, Fl, US.

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Studsvik EcoSafeStudsvik Eco & Safety AB S-611 82 NYKÖPING Sweden

Phone +46 155 22 16 00 Telefax +46 155 22 16 16

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Studsvik Report STUDSVIK ES-95 10

ADVANCED NUCLEAR REACTOR TYPESAND TECHNOLOGIES



Studsvik EcoSafe


1995-02-13


ADVANCED NUCLEAR REACTOR TYPES AND TECHNOLOGIES


Abstract

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fissionreactor types and fuel cycle technologies. Two pans have been prepared:

Pan IPartn

Reactors for Power ProductionHeating and Other Reactor Applications

Part HI, which will cover advanced waste management technology, reprocessing and disposal fordifferent nuclear fission options is planned for compilation during 1995.

This catalogue was prepared according to a special format which briefly presents the project title,technical approach, development status, application of the technology, reactor type, po\ 'er output,and organization which developed these designs.

Parts I and II cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and moltensalt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applicationsas power production, heat generation, ship propulsion, space power sources and transmutation of suchwaste are included.

Each project is described within a few pages with the main features of an actual design using a tablewith main technical data and figure as well as references for additional information. Each chapterstarts with an introduction which briefly describes main trends and approaches in this field.Explanations of terms and abbreviations are provided in a glossary.

ISBN 91-7010-258-9

© Studsvik Eco & Safety AB, Sweden 1995

Approved by

nA.


1995-02-13

Preface

For several years Studsvik AB has had the special task, disignated by theSwedish Ministry of Industry and Commerce of surveying, analysing andreporting on the development efforts and achievements abroad, concerningadvanced nuclear fission technology. Program management and most of thetask is performed within the subsidiary company Studsvik Eco & Safety AB.The results are presented in an annual summary report as well as in topicalreports in Swedish. Contacts and collaboration with colleagues abroad andother institutes and research organisations are of importance in this work. Inrecent years, close contacts have been established between the RussianResearch Centre "Kurchatov Institute", in Moscow and Studsvik. As a jointproject and in close co-operation it was decided to prepare a documentwhich briefly presented the plans, projects and achievements within the areaof advanced nuclear fission technology and their present status in variouscountries.

The first and second pans of this work concern nuclear fission reactors. Thelast will examine front- and back-end of advanced nuclear fuel cycles. Theintention is to extend and also update the material compiled in new editionswhen necessary. The first two parts were compiled by Victor Ignatiev(Editor), Olga Feinberg and Alexei Morozov who are leading scientists ofthe RRC "Kurchatov Institute" and I would like to thank them all for theirefforts which made this report possible.

Acknowledgements are due to Acad. N Ponomarev-Stepnoi andProf. A Gagarinski from RRC "Kurchatov Institute" who supported thework and J Kupitz, V Krett, C Goetzman from IAEA, Stanislav Subbotinefrom RCC "Kurchatov Institute", Tor Pedersen of ABB Atom and BengtPerehagen for providing necessary information and useful suggestions.

I would also like to thank other contributors, among them ElisabetAppelgren for the typing and Monica Bowen-Schrire for correcting theEnglish.

I hope that the report will be of help to those who would like to obtain anoverview of recent advancements in this area

Lennart DevellProgram ManagerAdvanced Nuclear Technology

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Table of Contents

Introduction

Table of Advanced Reactors

Description Format


Introduction

A Advanced Water-Cooled Reactors


AT

AEOAEP1AEP2AEP3AEP4AEP5AEP6AEP7

AEB1AEB2

AEH1

IntroducoonN4SizewellBKWU-ConvoyAPWRM/WSystem 80 PlusW E R 1000 (V-392)EPR

ABWRBWR90

CANDU9

Evolutionary designs for medium power

ATOATP1ATP2ATP3ATP4

ATB1ATB2ATB3ATB4ATB5

ATG1

ATH1

IntroductionW E R 500/600AP-600MS-600AC-600

SBWRHSBWRTOSBWR-900P1000 Natural Circulation BWRSWR-1000

MKER-800

CANDU3

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AI Innovative designs

AIO IntroductionAIP1 PIUS-600AIP2 ISERAIP3 ISISAIP4 MAPAIP5 VPBER-600AIP6 SIRAIP7 SPWRAIP8 B-500SKDIAIP9 SCLWR

B Liquid Metal Fast Reactors

BO IntroductionBTS1 EFRBTS2 BN600MBIS3 DEMOFBRBIS4 SAFRBIS5 ALMR (PRISM)BIL1 BREST 300BIL2 LFBR

Gas-Cooled Reactors

COCI1CI2CI3CI4

IntroductionHTR ModulMHTGRGT-MHRHTGR MHD

D Molten Sal; Reactors

DODI1DI2DI3

IntroductionUSRFUJI-PuMSR-NC

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1995-02-13


Introduction

E

F

G

i

District-Heating Reactors

EOElE2E3E4E5E6H7E8E9E!0Eli

IntroductionAST-500MMARSSECURE-HNHP-200Thermos-100NHR-200RUTAKNDHRSlowpoke (SES-10)GeyserGHR20

Decentralized Nuclear Heating

F0FlF2F3F4F5F6

Ship

GOGlG2G3G4

IntroductionCAREM-25PAES-100TRIGA Power SystemABVELENACompact HTGR Gas Turbine

Reactors

IntroductionOK-900AKLT-40MRXDRX

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Space Nuclear Reactors

10111213141516171819

IntroductionGPHS-RTGROMASHKATOPAZ-IISTAR-CSP-100ERATOLMCPBRRovcr/NERVANPPS

Subcritical Accelerator-Driven Waste Transmutationand Energy Generation Systems

JOJlJ2J3J4J5J6

IntroductionEnergy AmplifierPHOENIXJAERI-TPCATW/ABCAMSBBBR

Appendix 1 List of Abbreviations and Glossary of Terms

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1995-02-13

Introduction

Since the beginning of the nuclear fission era, illustrated e.g. by the start-up ofthe Fermi pile in 1942, increasing efforts have been made to develop and useimproved nuclear fission reactors. The civilian application of nuclear fission ismainly power plants for electricity generation but, to a small extent, also forheat generation. There are reactors for research, radio-nuclidc production,medical treatment, water desalination, ship propulsion and space applications.Also, some different systems in critical or subcritical accelerator driven modehave been studied for waste transmutation, weapon plutonium incineration,tritium production etc.

Many overviews and proceedings from recent conferences are available whichdescribe the evolution of reactor design, technologies and applications invarious countries (some of them are listed below). The present document is acomprehensive collection of short descriptions of new and advanced nuclearreactor types and technologies on a global scale (see Table 1). The purpose ofthe document is to provide an overview of present plans, projects, designs andconcepts of future reactors. The material has been collected from the openliterature and from information from the vendors.

The description of each design or concept follows a standardized format (seeTable 2) to facilitate comparison. The current status of the technology isdescribed for each design as well as type of application, reactor type, poweroutput etc.

Compilation of data began in January 1994 and the present document has beenissued as the first edition which is intended to be updated. The authors wouldtherefore appreciate any comments, corrections and additions.

References

Status of Advanced Technology and Design for Water-CooledReactors: Light Water ReactorsVienna, 1988, IAEA-TECDOC-479.

KABANOV, L, KUPITZ, J and GOETZMANN, C AAdvanced reactors: Safety and environmental considerations.An international perspective on the next generation of nuclearplants.IAEA Bulletin, 2/1992.

GAGARINSKI, A Yu, IGNATIEV, V V, NOVIKOV, V M andSUBBOTINE, S AAdvanced light-water reactor: Russian approachesIAEA Bulletin, 2/1992.

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1995-02-13

PERSHAGEN, BAdvanced Light Water Reactors (in Swedish)Studsvik AB, Sweden 1990 (STUDSVIK/NS-90/141).

PERSHAGEN, BNew Light Water Reactors (in Swedish)Studsvik AB, Sweden 1992 (STUDSVK/NS-92/61).

Small and medium reactors. Status and prospects. Report by anexpert group, OECD NEA, Paris, 1991.

ALEKSEEV, P N, IGNATIEV, V V, SUBBOTINE, S A e aReactor plant designs with the enhanced safety: analysis of theconcept (in Russian)Energoatemizdat, 1993.

FORSBERG, C W and REICH, W JWorld-wide Advanced Nuclear Power Reactors with Passive andInherent Safety; What, Why, How, and Who.Report ORNL/TM-11907, Sept, 1991.

ANGELO, J ASpace nuclear power, 1985, Orbit book company, USA.

ANS Topical Meeting on Safety of Next Generation PowerReactorsSeattle, May 1988.

IAEA Workshop on the Safety of Nuclear Installations: FutureDirectionsChicago, August 1989.

IAEA Technical Committee Meeting on Technical and EconomicAspects of High ConvertersNumberg, March 1990.

ENS/ANS-Foratom ENC90 ConferenceLyon, September 1990.

ENS TOPNUXr93 International ENS TOPical Meeting Towardsthe Next Generation of Light Water Reactors.The Hague 25 - 28 April, 1993.

Global '93 - International Conference and Technology Exhibitionon Future Nuclear Systems: Emerging Fuel Cycles and WasteDisposal Options.September 12 - 17,1993,, Seattle, Washington.

ww victor/Il-95-10 ea

STUDSVIK ECO & SAFETY STUDSVIKyES-95/10

1995-02-13

o

o

7th International Conference on Emerging Nuclear EnergySystems (ICENES '93).September 20 - 24,1993, Makuhari, Chiba, Japan.

Advanced nuclear power systems - design technology andstrategies for their deployment18-22 October 1993, Seoul, Rubublic of Korea.

ARS '94 International Topical Meeting on Advanced ReactorsSafety.17 - 21 April 1994, Pittsburg, USA.

Overview of physics aspects of Different Transmutation Concepts.1994, OECD, Report NEA/NSC/DOC(94)11.

Developing Technology to Reduce Radioactive Waste May TakeDecades and Be Costly.1993, December, USA, Report GAO/RCED-94-16.

AIP Conference Proceedings 324.1995, January 8-12, USA, NM, Albuquerque.

ww victor/II-95-10ea


1995-02-13

Table 1 Advanced Reactors


Abbre-viation

A

AE

AEP1

AEP2

AEP3

AEP4

AEP5

AEP6

AEP7

AEB1

AEB2

AEH1

AT

ATP1

ATP2

ATP3

ATP4

ATB1

ATB2

ATB3

Name Type Size/MW(e)

Country VendorOrganization

Developmentstatus

Advanced Water Cooled Reactors

Evolutionary Designs for Large Power Plants

N4

Sizewell B

KWU-Convoy

APWR

System 80 Plus

WER-1000(V-392)

EPR

ABWR

BWR90

CANDU9

PWR

PWR

PWR

PWR

PWR

PWR

PWR

BWR

BWR

HWR

1400

1250

1287

1250

1345

1000

1500

1356

1379

1050

France

US/UK

Germany

Japan/US

US

RF

Germany/France

Japan/US

Sweden

Canada

Framatome

Westinghouse/Nuclear Electric

Siemens-KWU

Miisubishi//KEPCO/Westinghouse

ABB-CE

Hydropress

Siemens-KWU/Framatome

Hitachi/Toshiba/GE

ABB Atom

AECL

Commercial

Commercial

Commercial

Detailed design

Detailed design

Detailed design

Basic design

Commercial

Commercial

Commercial

Evolutionary Designs for Medium Power Plants

VVER-50O/6OO

AP-600

MS-600

AC-600

SBWR

HSSWR

TOSBWR-900P

PWR

PWR

PWR

PWR

BWR

BWR

BWR

600

630

630

600

600

600

310

RF

US

Japan

China

US

Japan

Japan

Hydropress

Westinghouse

Mitsubishi

CNNC

General Electric

Hitachi

Toshiba

Detailed design

Detailed design

Basic design

Basic design

Detailed design

Basic design

Ba.' IC design

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1995-02-13


Abbre-viation

ATB4

ATB5

ATG1

ATH1

AI

AIP1

AIP2

AIP3

AIP4

A1P5

A1P6

AIP7

AIP8

AIP9

B

BT

BTS1

BTS2

BI

BIS3

BIS4

BIS5

Name

1000 NaturalCirculation BWR

SWR1000

MKER-800

CANDU3

Type

BWR

BWR

LWGR

HWR

Size/MW(e)

1000

1000

800

450

Country

US

Germany

RF

Canada

VendorOrganization

General Electric

Siemens-KWU

RDIPE

AECL

Developmentstatus

Studies

Basic design

Studies

Detailed design

Innovative Designs

PIUS-600

ISER

ISIS

MAP

VPBER-600

SIR

SPWR

B-500 SKD1

SCLWR

PWR

PWR

PWR

PWR

PWR

PWR

PWR

PWR

PWR

600

200

200

300

600

320

600

515

1100

Sweden

Japan

Italy

US

RF

US/UK

Japan

RF

Japan

ABB Atom

JAERI

Ansaldo Spa

CombustionEngineering

OKBM

ABB-CE/Rolls Royce

JAERI

RRC-KI/Hydropress

University of Tokyo

Basic design

Studies

Studies

Basic design

Basic design

Basic design

Studies

Studies

Studies

Liquid Metal Fast Reactors

Evolutionary Designs

EFR

BN-600M

LMR

LMR

1500

600

France/GB/Germany/Italy

RF

EdF/NE/Bayemwerk/ENEL

OKBM

Basic design

Detailed design

Innovative Designs

DEMOFBR

SAFR

ALMR (PRISM)

LMR

LMR

LMR

660

450

150

Japan

US

US

FEPS

Rockwell Int/CE

General Electric

Basic design

Basic design

Basic design

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STUDSV1K ECO & SAFETY STUDSVIK/ES-95/10

1995-02-13


Abbre-viation

BIL1

BIL2

C

CI1

CI2

CI3

CI4

D

DI1

DI2

DI3

Name

BREST 300

LFBR

Type

LMR

LMR

Size/MW(e)

300

625

Country

RF

Japan

VendorOrganization

RDIPE

JAERI

Developmentstatus

Studies

Studies

Gas Cooled Reactors/Innovative designs

HTR-Modul

MHTGR

GT-MHR

HTGR-MHD

HTGR

HTGR

HTGR

HTGR

80

190

300

860

Germany

US

US

Japan

Siemens-KWU

General Atomics

General Atomics

JAERI

Basic design

Basic design

Studies

Studies

Molten Salt Reactors/Innovative designs

USR

FUJl-Pu

MSR-NC

MSR

MSR

MSR

625

100

470

US

Japan

RF

ORNL

Tokai University

RRC-Ki

Studies

Studies

Studies

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1995-02-13

Table 1 Advanced Reactors (cont d)

Part D Heating and Other Reactor Applications

Abbre-viation

E

El

E2

E3

E4

E5

E6

E"7

E8

E9

E10

Ell

Name Type Size/MW(t)

Country VendorOrganization

Developmentstatus

District Heating Reactors

AST-500 M

MARS

SECURE-H

NHP-200

Thermos-100

NHR-200

RUTA

KNDHR

Slowpoke (SES-10)

Geyser

GHR-20

PWR

PWR

PWR

PWR

PWR

PWR

LWR

LWR

LWR

PWR

HTGR

500

600

400

200

100

200

20

10

10

10-50

20

RF

Italy

Sweden

Germany

France

China

RF

Korea

Canada

Switzerland

Germany/Switz.

OKBM

Rome University

ASEA

Siemens-KWU

CEA

INET

RDIPE

KAERI

AECL

PSI

KWU

Commerciaj

Studies

Studies

Detailed design

Basic design

Detailed design

Studies

Studies

Commercial

Studies

Basic design

F Decentralized Nuclear Heating Power Reactors

Fl

F2

F3

F4

F5

F6

G

Gl

G2

CAREM-25

PAES-100

TR1GA Power System

ABV

ELENA

Compact HTGR GasTurbine

PWR

PWR

PWR

PWR

PWR

HTGR

100

160-170

64

60

3

29

Argentina

RF

US

RF

RF

US

INVAP

OKBM

General Atomic

OKBM

RRC-KI

General Atomic

Studies

Commercial

Commercial

Detailed design

Basic design

Studies

Ship Reactors

OK-900A

KLT-JO

PWR

PWR

170

170

RF

RF

OKBM

OKBM

Commercial

Commercial



1995-02-13

Table

Abbre-viation

G3

G4

I

11

12

13

14

15

16

17

18

19

1 Advanced Reactors (cont'd)

Name

MRX

DRX

Type

PWR

PWR

Size/MW(t)

100

0.75

Country

Japan

Japan

VendorOrganization

JAER1

JAERI

Develop rtientstatus

Basic design

Basic design

Space Nuclear Reactors and Isotope Batteries

GPHS-RTG

ROMASHKA

TOPAZ-U

STAR-C

SP-100

ERATO

LMCPBR

Rover/NERVA

NPPS

RadioisotopeThermo-electric

Fast,Thermo-electric

LMR,Thermionic

Thermionic

Fast,Thermionic

LMR/HTR

LMR/HTR

HTGR

HTGR

3 104MWe

3-10-MO-2

MWe

610-3MWe

510-3-410-2

MWe

0.1 MWe

2 10-2MWe

10

367 - 1 566

1200

US

RF

RF

US

US

France

Japan

US

RF

Westinghouse

RRC-KI

RRC-KI

Rockwell Institute

General Electric

CNES/CEA

JAERI, Toshiba

LANL,Westinghouse

RRC-KI

Commercial

Detaileddesign

Commerical

Basic design

Basic design

Studies

Studies

Basic design

Studies

J Subcritbal Accelerator Driven Waste Transmutation and Energy Generation Systems

Jl

J2

J3

)4

J5

J6

Energy Amplifier

PHOENIX

JAERI-TPC

ATW/ABC

AMSB

BBR

LWR/HTGR

LMR

MSR/LMR

MSR/LMR

MSR

MSR

200

3600

820

3000

1 000-2 100

5000

Switzerland

US

Japan

US

Japan

France

CERN

BNL

JAERI

LANL

Tokay University

CEA

Studies

Studies

Studies

Studies

Studies

Studies


STUDSVIK ECO & SAFETY STUDSVlK/ES-95/10

1995-02-13

Table 2 Advanced Nuclear Design Description Format

Format Description

Technical approach

Tide

Applicaton of the reactor

ETI

Title o

NPPNHPNHPPSMSSNSTWS

Reactor type

Power output

Organization

Development status

Description

Schedule

References

Evolutionary design for large power plantsEvolutionary design for medium power plantsInnovative design

Nuclear power plantNuclear heating plantNuclear heat and power plantShip mobile systemSpace nuclear sourceTransmutation waste system

PWR, BWR, HWR, LWGR, FBR, HTGR, MSR, SADS(subcritical accelerator-driven system) e.a.

L Large power (more than 1 000 MWe)M Medium power (more than 100 MWe, less than 1 000 MWe)S Small power (more than 1 MWe, less than 100 MWe)VS Very small power (less than 1 MWe)

Type of organization, which developed design

C Commercially available design

D Detailed engineering design

B Basic engineering design

S Studies of concept

General description of the design. This will generally include asimple description of the reactor plant. This may be followed by asummary with the main features of an actual design using tableswith main technical data or figures

Schedule for development

References for additional information.

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E

E0

District Heating Reactors

Introduction

The potential for the nuclear heating application depends on the specifictemperature requirements of different technologies provided below:

Networks

Application

District heating and seadesalination

Petroleum refining

Petrochemical industry

Hydrogen production

Requiredtemperature,°C

80-200

250 - 550

600-880

900 - 1 000

Nuclear source

Reactortype

LWR

LMR

MSR

HTGR

Supplytemperature,°C

300

540

700

950

The current nuclear application is mainly considered for district heatingdistribution, but to a lesser extent also for industrial one (e.g. MARS reactorplant (see E2) and HTR-Modul (see CI1)).

District heating markets are currently restricted to specific northerncountries, such as Finland, Gennany, Sweden, Canada and Russia where heatdistribution networks are already in existence in certain regions. More than10 different NHP concepts are known today throughout the world. Most ofthem are based on LWR technology and designed to satisfy different energyboundary conditions and have following different technological lines.

The specific features of NHP are the following:

o The power ranges of NHP (from 2 to 500 MWt) are con-siderably lower than those of current and advanced LWRs

o The supply temperature of the district-heating networks fed bythese reactors does not exceed 150 °C; the coolant temperaturein the primary circuit (e.g. about 200 °C for water-cooledNHP) is therefore considerably lower than for NPPs (thetypical temperature for PWR is 320 °C and for BWR 300 °C)

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E:2(2)

o The power density of most NHP with PWR reactor plant lies inthe range of not more than 55 kW/l compared with about100 kW/l for large power and 60 - 70 kW/l for medium powerAPWR

o The low temperature required allows for lower operatingpressures, which in turn, along with the smaller size, lead to lessmassive RPVs and allows the components of the primary circuitto be integrated within the RPV

o All the NHPs considered make as much use as possible of com-ponents and systems proven in the operation of large NPPs

o The operation schemes proposed for all NHPs are considerablysimplified as compared to large NPPs. Some of them canoperate without operation staff

o The time between fuelling in NHPs is in general longer than inNPPs. Almost all concepts are based on existing fuel cycletechnologies

o Most of the NHP concepts considered make extensive use ofpassive systems and components and rely more heavily onnatural processes rather than engineered safeguards

However, only some reacton for district heating: AST-500 in Russia,SLOWPOKE in Canada and NHP in China have actually been realized so farbut no operation experience of advanced designs has been gained as yet

References

o

SEMENOV, B A, KUPITZ, J and CLEVELAND, JProspects for development of advanced reactors.Proceedings International Conference on Design and Safety ofAdvanced Nuclear Power Plants, 25-29 Oct 1992, Tongo,Japan, Vol I,p2.6.

Small and medium reactors.Technical supplementNuclear Energy Agency, OECD, Paris, 1991, Vol 1,2.

GAGARINSKI, A Yu et alAdvanced LWRs: Analysis of new approaches and ideas.NSI, Moscow, 1993.

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El

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Schedule

Description

AST-500M

NHP

PWR

Medium power/500 MWt

OKBM.RF

Commercial

The AST-500M district heating plant is a PWR with an integral arrangementof the primary circuit (core and heat-exchangers are located in one vessel)and natural circulation of coolant in the first circuits. The second circuit has apressurizer. About 20 % of the nominal power can be removed under naturalcirculation conditions in the second circuit The capacity of rhe nominalresidual sink is equal to 2.5% of the nominal reactor power. The reactor hasthree loops for nominal heat removal and three independent emergency heatremoval systems (EHRS). Each EHRS has a water storage tank connected tothe loops of the second circuit, and can remove 2 % of nominal power. Thequantity of water is enough for three days of cooling. For the emergencymake up of the reactor during the AST-500M reactor plant black-out in thedepressurized state (in particular during refuelling operations the EHRS isnot capable of ensuring cool-down because of elimination of circulation overthe second circuit) three independent pipes for the reactor make up with thewater flowing by gravity from EHRS tanks are used. The containment(Voronej AST-500 plant) and shield building (Nijegorodskaj AST-500 plant)is used for protection from external events. The guard vessel is used for thereduction of loss of coolant accident probability. The calculated core damageprobability is less than 10'7 per reactor-year. There is also a smaller unit,AST-200 district heating plant, of 200 MWt output.

Two AST-500M units intended for district heating are under constructionnear Voronej, Russia.

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1995-02-13

El:2(3)

AST-500M Unit Data

Reactor power 500 MWt

Parameters of the first circuitPressureInlet coolant temperatureOutlet coolant temperature

Parameters of the intermediate circuitPressureTemperature in the heat-exchanger.inlet/outlet

20 bar131 °C208 °C

12 bar

87/160 °C

Parameters of district-heating water (inside plant boundary)Pressure 20 barTemperature, inlet/outlet 64/144 °C

Core parameters:DiameterHeightPower densityFuel loadingFuel residence timeRefuellingsFuel burnup

Reactor vesselDiameter/heightWater volumeNeutron fluence

Guard vesselDiameter/height

2.8 m3.0 m30 MW/m3

50.0 t6 years314MWd/kg

5.3/25.3 m190 m3

1016 n/cm2

8.15/27.95 m

References MITENKOV, F M et alAdvanced reactor plant of passive safety for district heating plantsAST-500M.Proceedings International Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 25-29,1992, Tokyo, Japan, Vol 1, p 4.1.

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1995-02-13

El:3(3)

1 Reactor2 Guard vessel3 Boron solution storage tank4 ERHR condenser5 Water storage tank6 Bubbler7 Containment8 Secondary circuit steam pressurizer9 EHRH heat exchanger with water storage tank

10 Heating grid HX11 Secondary coolant clean-up system12 Primary coolant clean-up system13 Secondary circuit make-up14 Makeup by water, boron solution feed15 Make-up by water, gas

J L

Figure ElAST-500M: Flow diagram.

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1995-02-13

E2:i(4)

E2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

MARS (Multipurpose Advanced Reactor Inherently Safe)

NHP.NHPP

PWR


Rome University, Italy

Studies

The MARS design concept combines the advantages of s«PWR (separation of the reactor coolant from the consumer workingmedium), BWR (lower coolant pressure and pools with a large cold waterinventory within the containment), AST district heating plant (systems ofemergency residual heat removal through the intermediate circuit withnatural circulation), and HTGR (prestressed reinforced concrete guardvessel).

The MARS reactor safety is improved primarily due to reduced stresses inthe elements of the primary circuit structures as all systems of the primarycircuit are accommodated within a cavity of prestressed reinforced concrete,filled with cold water at a pressure equal to that of the primary circuit. Thispractically eliminates the accidents associated with depressurization andruptures of the pipes in the primary circuit and reactor vessel.

The reactor is supposed to be used for industrial heating. For increasing thereliability and safety of the main technological process it was consideredeconomically favourable to use the following technical solutions:

o Low average core power density (63 MW/m^) and medium unitpower (600 MWt) permit removal of the decay heat byemploying natural circulation of the coolant in all circuits

o Low pressure reactor coolant (7 MPa), 40 °C coolanttr nperature margin to the saturation temperature at the reactoroutlet (246 °C) and moderate maximum heat fluxes ensure aconsiderable critical heat flux ratio in transient and accidentsituations

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E2:2(4)

Schedule

o Stainless steel fuel cladding is resistant to the chemicalreactions with the coolant, reduces the probability of fissionproduct escaping into the coolant and, thus, reduces theoccupational exposure and simplifies reactor maintenance.

The reactivity control is accomplished using the standard PWR clusters ofabsorbing rods. For the reactor scram the mechanism of passive insertion ofthe control rods at the core temperature exceeding the specified limit hasbeen developed. In accordance with industrial requirements the time ofcontinuous operation between the refuellings has been established as17 months, which requires the use of liquid boron control of reactivity (themaximum boron content in the beginning of the fuel cycle is 1200 p.p.m.).

The MARS concept uses flange joints, instead of weld ones, connecting thepipelines with the reactor vessel and steam generators.

The inner containment sections are designed as a system of six inter-connected cylinders of prestressed reinforced concrete. In the upper part ofthe outer containment, the pools of two independent emergency cool-downsystems are located. These pools are also used for spent fuel storage.

Innovative, passive technologies for check valve design and the control rodscram system have been used.

No MARS plant has been ordered yet The conceptual study has beencompleted. No current development work is being carried out.

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1995-02-13

E2:3(4)

MARS Unit Data

Net plant capability 600 MWt

Coolant temperatureInletOutlet

Coolant pressure

Outer diameter of the fuel rod

Fuel lattice pitch

Core height

Fuel assembly array



Average power density of the core

Average burnup

Fuel enrichment (first loading)

Fuel residence time

216 °C246 °C

70 bar

9.78 mm

13.0 mm

260 cm

15x15

96

216 cm

63 MW/m3

30MWd/kg

3.6%

51 months

References Safety aspects of designs for future light water reactors (evolutionaryreactors)IAEA-TECDOC-712, IAEA, Vienna, 1993.

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STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 E2:4(4)

•;•;:.:• •:"••'•:.••: -v::- •;'••:• /•:..: r ; f . b n v . - / - . ^ : . : - : : , : - : r « : . ; , .

Figure E2MARS: Reactor building.

ReactorSteam generatorCheck valveMCP

5 Inner containment6 Pool7 Bride crane8 Refuelling machine

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1995-02-13

E3:l(3)

E3

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

SECURE-H

NHP

PWR


ASEA, Sweden

Studies

SECURE-H reactor plant

Schedule

on well-known PIUS principle (see AIP1). The reactor core is located at thebottom of a prestressed concrete vessel, which contains some 1 500 m^ ofcold heavily-borated water. The pressure in this vessel is 7 bar. The reactortank separates the heavily-borated water of the pool from the less heavilyborated water of the primary circuit The SECURE-H plant has a two-looparrangement of the primary circuit. The primary pumps and heat-exchangersare located outside the prestressed concrete vessel. The cooling water isheated up by the core from 90 °C to 120 °C. Reactivity control is achievedby the adjustment of the boron content of the primary circuit. The corepower density is low, which means good margins against fuel failure.

The particular safety feature of the SECURE-H reactor plant is its primarycircuit, which is open towards the borated water of the pool. During normaloperation, a pressure equilibrium is established between primary water andpool water. The pressure drop over the core is only 0.11 bar. In case ofdisturbance of the equilibrium, after a pump failure or due to a temperaturerise in the primary circuit, the borated water of the pool enters the primarycircuit and shuts down the reactor.

The primary cooling system and the blowdown chamber are housed in acontainment building below ground. The primary cooling system deli versheat from the core through the primary heat-exchanger to an intermediatecooling system connected to the district-heating grid through a secondaryheat-exchanger. The intermediate cooling system operates at a higherpressure than the primary system, and its water has a high content of boron.

No SECURE-H reactor plant has been ordered yet. No current developmentworkis being carried out.

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STUDSVIK ECO & SAFETY AB•TY AB STUDSVIK/ES-95/10

1995-02-13

SECURE-H Unit Date

Thermal output

Fuel power density


Number of fuel rod positions per assembly

Active core height

Equivalent core diameter

Fuel enrichment

Core coolant flow

Primary system operating pressure

Coolant inlet temperature

Coolant outlet temperature

Power density

Inner diameter of concrete RPV

Fuel residence time

400 MWt

15.0kW/kgU

308

8x8

1.845 m

2.51m

2%

2 300kg/s

20 bar

150 °C

190 °C

45kW/l

9.5 m

2.5 years

E3:2(3)

References PIND.CThe SECURE Heating Reactor.Nuclear Technology, Vol 79, Nov, 1987.

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STUDSVIK ECO & SAFETY AB STUDTVIK/ES-95/10

1995-02-13

E3:3(3)

District heoting grd

Intermediote cooling circuit

Pnmory cooling circuit

Slowdown chamber

Reactor core

Concrete vessel

Pool water

Figure E3SECURE-H: Main cooling system.

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1995-02-13

E4:i(3)

E4

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

NHP-200

NHP

FWR

Small power/200 MWt


Detailed design

The NHP-200 reactor coi

Schedule

BWR design for nuclear power plants. The plant size has been fixed at 200M W L The primary circuit cooling water enters the core at a pressure of 15bar and a temperature of 160 °C. The core configuration is similar to a BWRplant design.

The temperature of the water/steam mixture at the core outlet is 200 °C. Therather low pressure of the primary coolant and low pressure capacity leadsquite naturally to an integration of the primary heat-exchangers, the hydraulicdrives of the control rods and the storage racks for spent fuel within thepressure vessel. The core heat is transferred to twelve heat-exchangers bynatural circulation. The reactor pressure vessel is surrounded by a tight-fitting containment. A particular safety feature is that the core cannot beuncovered in case of a leakage in the primary pressure barrier, due to thesmall volume available between the RPV and the containment Reactorshutdown is achieved by switching off the pumps of the hydraulic control roddrives. Residual heat removal is achieved by opening the valves to coolingcircuits with natural circulation cooling towers.

The intermediate circuit has a higher pressure compared to the primarysystem. The entire heating plant is housed in a single building complex withthree parts: a reactor building, which is embedded in the ground, a reactorbuilding operating hall and an auxiliary building.

No NHP-200 plant has been ordered yet. The main design principles arealready demonstrated in a prototype heating reactor in China.

ww victor/e4 ea

STUDSVnC ECO & SAFETY AB•TY AB STUDSVIK/ES-95/10

1995-02-13

NHP-200 Unit Data

Thermal power

Core coolant flow


Fuel assembly type

Active core height

Primary pressure

Coolant temperature

Average volumetric steam contentat core outlet

Maximum heat flux


Average specific power


Network temperature (forward/return)

E4:2(

200 MWt

1030kg/s

180

8x8, BWR type

2.350 m

15 bar

158/198 °C

26%

670kW/m2

20kW/l

10kW/kg

45

120/80 °C

References Small and medium reactors. Technical supplement,Nuclear Energy Agency. OECD, Paris, France, 1991, Vol 2, p 193-198.

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1995-02-13

E4:3(3)

Intermediatecircuit

Chimney

Spentfuel rackCore

PressurevesselBiologicalshield

Figure E4NHP-200: Primary system arrangement.

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1995-02-13

E5:i(2)

E5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

THERMOS-100

NHP

PWR

Small power/100 MWt

CEA, France

Basic design

Description

Schedule

The French Commissariat å l'Energie Atomique (CEA) has been developingthe Thermos reactor plant since the beginning of the 1970s. It is a PWR,designed for a thermal output of 100 MW. The reactor core is located in astainless-steel vessel filled with water. This vessel is, in tum, located in areinforced concrete pool full of borated water. The reactor vessel containsthree coolant pumps. The fuel elements are of the slab-type, with Zircaloycladding. This fuel type has been developed for the power reactors of sub-marines.

No THERMOS reactor plant has been ordered yet No current developmentwork is being carried out.

ww victor/e5 ea


1995-02-13

E5:2(2)

THERMOS-100 Unit Data

References

Power range

Power density

Primary circuit type

Primary coolant


Primary coolant temperature, inlet/outlet

Network forward temperature

Heat removal by

Residual heat removal by

Ultimate heat sink

Fuel type

Fuel residence tune

Fuel enrichment

Reactivity control by

Shutdown by

Small and medium reactors. Technical suDDlem

100 MWt

45kW/l

integrated

H2O

11 bar

131/144 °C

130 °C

pump

natural circulation

pool

UO2 slabs

2 years

3.7

rods

reds

entNuclear Energy Agency. OECD, Paris, France, 1991, Vol 1,2.

ww viclor/e5 ca

STUDS VIK ECO & SAFETY r>v STUDSVnC/ES-95/10

1995-02-13

E6:i(3)

E6

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

NHR-200

NHP

PWR

Small power/2C

INET (Institute

Detailed design

In 1989 the 5M

Schedule

In 1989 the 5MW Test Heating Reactor was put into operation in China. In1989 it was decided to construct a commercial NHR-200 with an output of200 MWt The main design principles of the NHR-200 are similar to theNHP-200 Siemens design.

The main technical features of the NHR-200 are as follows:

o

o

o

o

o

o

o

o

Integral arrangement

Natural circulation

Double pressure vessel

Low temperature, low pressure, low power density

Hydraulic facility of control rods

Storage of spent fuel in the pressure vessel around the reactorcore

No emergency cooling system

Residual heat removal system also uses natural circulation ofcoolant

The core is located at the bottom of the RPV. The primary heat-exchangersare arranged on the periphery in the upper part of the RPV. The systempressure is maintained by inert gas and steam. A containment fits tightlyaround the RPV so that the core will not become uncovered under anypostulated coolant leaks.

It was decided to build a demonstration NHR-200 in the North of China in1995, which was classified as one of the key national nuclear projects. Thedetailed design study for the NHR-200 plant is finished. The plant is plannedto be completed in 1998.

ww victor/e6 ea


1995-02-13

E6:2(3)

NHR-200 Unit Data

Total thermal power

Reactor pressure


Core inlet temperature

Height of the core

Number of fuel bundles in core

Number of rods per bundle

Load of UO2

Enrichment of the fuel

Diameter of the fuel rod


Pressure of the secondary loop water

Temperature of the secondary loop water

Temperature of the heat grid water

200 MWt

22 bar

200 °C

138 °C

2.1m

120

141

-211

3.0, 2.4, 1.8

1.0 cm

-38 kW/l

26 bar

130/85 °C

110/70 °C

References CHANGWEN, M A and DAZHONG, WA safe and simple nuclear demonstration heating plant in China.Proceedings International Conference c Design and Safety of AdvancedNuclear Power Plants, Oct 25-29,1992, Tokyo, Japan, Vol 1, p 6-2.

DAZHONG, W and CHANGWEN, M AChina plans first 200 MW district heating reactor.Nuclear Engineering International 1994, Sept, p 60-62.

ww victor/e6 ea


1995-02-13

E6:3(3)

Figure E6NHR-200: Reactor building.

1 Primary heat exchanger2 Riser3 Biological shield

456

C1- "itainmentPressure vesselCore

ww victor/e6 ea

STUDSVK ECO & SAFETY AB STUDSVKyES-95/10

1995-02-13

E7:i(3)

E7

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

RUTA

NHP

LWR

Small powei

RDIPE (Res

Studies

The RUTA 1

Schedule

RDIPE (Research and Development Institute of Power Engineering), RF

The RUTA heating plant is a pool-type reactor with a 20 MWt output Allthe primary equipment is integrated within the cylindrical reactor tank. Thepool arrangement constitutes a circuit of natural circulation of the primarycoolant, whereby heat is transferred from the core to the primary heat-exchangers and further on to the heating network via an intermediate circuitand the network heat-exchangers.

The coolant circulation in the intermediate circuit is a natural one under allreactor conditions. The steel secondary circuit surfaces are used for decayheat removal from the reactor to the environment without reactor waterboiling during any accidents in network circuit One loop provides heatremoval from the reactor to the environment without water boiling in areactor when two others have failed.

In the event of severe external impact, when all intermediate circuits fail, theprovision is made for the a transfer from the reactor tank to the concretewell, then to the ground and for reactor water evaporation into the leak-tightreactor room (if it is necessary). This design forms a completely passivedecay heat removal system.

Underground siting of the reactor plant including secondary heat-exchangershas been proposed.

No RUTA reactor plant has been ordered yet The conceptual study has beencompleted.

ww victor/e7 ea

STUDSVK ECO & SAFETY ABiTY AB STUDSVK/ES-95/10

1995-02-13

RUTA Unit Data

Power

Coolant pressurePrimary circuitIntermediate circuitNetwork

Inlet/outlet coolant temperaturePrimary circuitIntermediate circuitNetwork

Outside tank diameter

Tank height

Number of loops in secondary circuit

Core dimensionsActive heightActive diameter


Fuel cycle

Average bumup

Enrichment

Reactor lifetime (with account of specialprocess for tank manufacturing)

E7:2(3)

20MWt

atmosphere4 bar6 - 1 bar

70/100 °C65/95 °C60/90 °C

4.8 m

15 m

3

1.0 m1.23 m

61

up to 1900 eff. day(8 yr)

23 MWd/kgU

4.0%

80-100 years

References ADAMOV, E O et alLow-power nuclear plants for district heating with pool-type reactor(RUTA).Proceedings International Conference on Design Safety and AdvancedNuclear Power Plants, Oct 25-29,1992, Tokyo, Japan, Vol 1, p 4.2.

ww victor/c7 ea


1995-02-13

E7:3(3)

Figure K7RUTA: Schematic.

1 Core, 2 Safeguard vessel, 3 Reactor tank, 4 Control rod, 5 RCP,6 Secondary heat exchanger, 7 Pipelines of secondary circuit,X Primary heat exchanger

ww viclor/e7 ca


1995-02-13

E8:i(3)

E8

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

KNDHR

NHP

LWR

Small powei

KAERI(Ko

Studies

KNDHR is;

Schedule

KAERI (Korean Atomic Energy Research Institute), Korea

KNDHR is a pool-type reactor with a maximum thermal power of 10 MWsimilar to SES-10 (Canada) and RUTA (Russia). It consists of three closedloops with natural circulation of the coolanL The heated water from the coretransfers to two plate-type heat-exchangers, situated in the upper part of thepool and the outcoming water from the heat exchangers returns to the core.The KNDHR reactor plant operates at atmospheric pressure. The secondarycircuit delivers the heat from the primary circuit to the building water supplysystem (user grid circuit) via the secondary heat-exchanger.

In KAERI, a feasibility study for KNDHR development in Korea has beenperformed in 1991. No KNDHR reactor plant has been ordered yet.

ww victor/e8 ea


1995-02-13

E8:2(3)

KNDHR Unit Data

Maximum thermal power

Reactor size

Core size

Secondary pump head

Secondary circuit size

Secondary circuit pressure

User-grid circuit pressure

Core inlet/outlet temperature

Secondary circuit inlet/outlet temperature

User-grid circuit inlet/outlet temperature

Core flow rate

Secondary circuit flow rate

User-grid circuit flow-rate

Reactor free surface pressure

lOMWt

05.3x13 m

01.1x1.2 m

10 m

00.5x37 m

3 bar

lbar

76.1/94.0 °C

69.7/89.7 °C

65.3/85.3 °C

122.3 kg/s

119.2 kg/s

120.6 kg/s

1.1 bar

References LEE, S I, KIM, D S and MOON, K SThermal hydraulic analysis on the Korean nuclear district heating reactor.Proceedings International Conference on Design and Safety of AdvanceNuclear Power Plants, Oct 25-29,1992, Tokyo, Japan, Vol 1, p 4.3.

ww victor/e8 ea


1995-02-13

E8:3(3)

Pump

Reactor Building

Return

Supply

i "i n

~i "i i

"i "i i

i n n

Core

Figure E8KNDHR. Schematic.

SecondaryHeat Exchanger

PrimaryHeat Exchanger

Reactor

Concrete Container

Ground Level

ww victor/e8 ea


1995-02-13

E9:i(3)

E9

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SES-10 (Slowpoke Energy System;

NHP

LWR

Small power/10 MWt

AECL (Atomic Energy of Canada Limited), Canada

Commercial

Based on the experience with its proven small 20 kV

Schedule

pool reactors in operation in half of the Canadian research centres, the AECLhas developed a swimming-pool reactor (SES-10) for heating purposes. Onedesign goal of the SES-10 reactor plant is to fully automate all essentialsystems. Thus the unit can be operated for extended periods without theoperator in the reactor building. The SES-10 plant has been designed for athermal output of 5 - 20 MW. Its core is situated in a metallic cylinder nearthe bottom of a water pool and consists of CANDU-proven fuel elementswith 3 per cent enriched uranium in a Zlrcaloy cladding. Natural circulationtransfers the heat from the core to the heat-exchangers located at the top.Atmospheric pressure prevails at the water surface. Core reactivity isregulated by means of a single central control rod. The long-term reactivitycompensation is achieved by means of a movable beryllium reflector. Thereactor is designed to be operated unmanned. A particular safety feature isthe lack of need for any pressure vessel, with the disadvantage that thesupply temperature in the heating network is limited to 85 °C.

The SES-10 reactor plant can heat 150 000 m2 of floor area or approxi-mately 1 500 individual apartments with an inflation-resistant fuel source.The reactor core and safety systems are being designed to remain fullyfunctional during and after seismic events. Use of factory-fabricated modulescontributes to the short construction time of one year proposed for thecommercial use.

A 2 MWt SDR demonstration plant of this type has been in successfuloperation in Whiteshell/Manitoba (Canada) since July 1987. No SES-10reactor plant has been ordered yet.

ww viklor/e9 ea


1995-02-13

E9:2(3)

SES-10 Unit Data

Power range

Power density


Primary coolant


Primary coolant temperature inlet/outlet

Network temperature forward/return

Heat removal by


Ultimate heat sink

Fuel type

Fuel residence time

Fuel enrichment


Shutdown by

2 - 10 MWt

5.5 kW/l

tank

H2O

1.7 bar

60/90 °C

85/70 °C

natural circulation

natural circulation

pool

UO2 rods

3 years

5%

Be-reflector

boric acid

References LYNCH, G Feta lUnattended Nuclear Systems for Local Energy Supply,Paper 4.2-2, Thirteenth Congress of the World Energy Conference, Cannes,France, 1986.

MARCHILDON, PRecent Developments in Canadian Nuclear Plant Licensing Practices.Proceedings Annual Conference of the Canadian Nuclear Society, Ottawa,Canada, 1985.

NATALIZIO, AUp-Front Licensing - A new Approach.Proceedings Annual Conference of the Canadian Nuclear Society, Ottawa,Canada, 1985.

ww viktor/e9 ea


1995-02-13

E9:3(3)

SLOWPOKE ENERGY SYSTEM

12 metres

WATER

HEATEXCHANGER

NATURAL.CIRCULATION

CORE

Figure E9SES-10: Schematic diagram.

ww viktor/e9 ea


1995-02-13

E10:i(3)

E10

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

Geyser

NHP

PWR

Small power/10 - 50 MWt

PSI (Paul Scherrer Institute), Switzerland

Studies

The design requirements set for the Geyser reac

o Low power density and sizeo Completely autonomous operationo Emphasis on simplicityo Underground location

Schedule

The Geyser reactor concept developed at the Paul Scherrer Institute (PSI)makes use of the static pressure of a high water column and does not need aproper pressure vessel. The reactor is located at the bottom of a concretewell some 50 m in depth and 5 m in diameter. With this arrangement thecoolant reaches a saturated state at 150 °C at the core outlet. During its risein a so-called diffuser, the coolant goes partially into the steam phase andtransfers its heat to the primary heat-exchangers, which are also located inthe well. These operate on their primary side as condensers and coolers andon the secondary side as evaporators. The secondary steam thus produced,transfers its heat through condensation to the heating network. Primary andsecondary circuits operate by natural circulation.

A particular safety feature is the large amount of borated water in the well,which, depending on the equilibrium between the heat produced in the coreand the heat demand by the network, can enter the primary circuit thusleading to a load dependent self-control and when necessary to a shutdown(see PIUS principle).

As another potential application sea water desalination is identified.

No Geyser plant has been ordered yet. The conceptual design study has beencompleted.

ww victor/elOea


1995-02-13

E10:2(3)

Geyser Unit Data

Power range

Power density


Primary coolant




Heat removal by


Ultimate heat sink

Fuel type

Fuel residence time

Fuel enrichment


Shutdown by

10-50 MWt

5kW/l

integrated

H2O

4.7 bar

135/149 °C

118/60 °C

natural circulation

natural circulation

pool

UO2 or UZrH rods

15 years

8-20

boric acid

boric acid

References Small and medium reactors. Status and prospects,Nuclear Energy Agency. OECD, Paris, France, 1991, Vol 1.

ww vie tor/el Oea


1995-02-13

E10:3(3)

Core ExchangDevice

Hoodln;CoiilMinc-c

PrimtryCondenser

Tray

.aypr Accumulator

, e r e

D 0 U u i

Figure E10Geyser: Layout

ww victor/elOea


1995-02-13

Ell

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

GHR-20

NHP

HTGR

Small power/20 MWt

KWU, Germany/Switzerland

Basic design

The GHR-20 district-heating

Schedule

system of 20 MWt for generating 80 kg/s pressurized hot water at 15 bar and120 °C with return temperature about 60 °C.

The GHR-20 reactor and the secondary system are designed for a service lifeof 40 years with a capacity factor of 0.5. The core structure and fuelelements are based on the spherical fuel element containing low enriched fueland on periodical refuelling. Reactivity control and shutdown are performedby absorber rods which are inserted into axial boreholes of the radialreflector. The primary circuit is designed with helium coolant inlet and outlettemperatures of 250 °C and 450 °C, with pressures of about 40 bar. The fuelelements have the standardized 6 cm diameter and contains TRISO-coatedparticles. Core meltdown is excluded due to the small power density of thefuel element and the small size of the reactor.

The GHR 20 plant is designed so that during operation the reactor can beoperated unmanned safely and with a high availability in a fully automaticway except for periodic inspections.

Ready for detailed design development. No current development work.

ww victor/e 11 ea

STUDSVIK ECO & SAFETY AB STUDS VK/ES-95/10

1995-02-13

E l 1:2(3)

GHR-20 Unit Data

Power range

Power density


Primary coolant




Heat removal by


Ultimate heat sink

Fuel type

Fuel residence time

Fuel enrichment


Shutdown by

10 - 50 MWt

2kW/l

integrated

helium

40 bar

250/450 »C

120/160 oC

pump

natural circulation

graphite

LEU-graphite pebbles

16 years

8

rods

rods

References Small and medium reactors. Technical Supplement,Nuclear Energy Agency. OECD, Paris, France, 1991, Vol II.

ww victor/ell ea


1995-02-13

Ell:3(3)

Prestressedconcretereoctor vessel

Graphite reflector

Reactor core

Heliumcirculator

Absorber rod

Thermal shield

- Liner withcooling pipes

Figure El lGHR-20: Reactor plant.

ww victor/ell ea


1995-02-13

F0

Decentralized Nuclear Heating PowerReactors

Introduction

The economic and geographical factors give a rise to a specific structure ofpower installations for electricity, drinking water production and heat supplyfor users located in hard-to-access isolated regions (e.g. far North and NorthEast in Russia and others). The creation of multi-purpose, small powernuclear plants for remote regions is one of the possible solutions.

Every one of these applications has been technically demonstrated. However,no commercial market has been established for any of these applications.This can be attributed to two reasons:

o The global problems encountered by any nuclear application areworld wide

o The availability of fossil fuel alternatives in these markets.Generally, these alternatives are considered cheaper and easierto manage in a general sense: technically, financially and fromthe project, regulatory and environmental points of views

The most essential conditions and requirements for decentralized nuclearpower heating plants (DNHPP) include the following:

o Energy sources virtually run under autonomous conditions,they have to control varying loads, and therefore need highmanoeuvrability

o Great distance away from the industrially developed regionsand centres of the country, the seasonal character andunreliability of transport communications result in rigidrequirements for energy source reliability, simplification of theiroperation and maintenance

o Severe climatic conditions mean rigid requirements for thesecurity of the consumers' power supply. In particular, itinvolves heat supply

The main part of DNHPP specific features is the following:

o Traditionally verified and proved technical solutions

o Natural circulation of primary coolant under all operatingconditions

ww victor/F-dnhpp ea


1995-02-13

F:2(2)

o

o

o

o

Natural coolant circulation in all circuits with a total loss ofpower or some other measures ensuring the passive principle ofdecay heat removal

Low power density of fuel

Large heat-accumulating capacity of reactor

Integral arrangement of the primary circuit

More than 10 different DNHPP concepts for remote earth regions in Russia,the USA, Japan and Argentina are known today throughout the world. Mostof them are based on LWR technology. Some of them represent a floatingbased design line. One HTGR concept is also under consideration. The unitcapacity is from 3 to 340 M W L

References

o SEKIMOTO, H (editor)Potential of small nuclear reactors for future clean and safesources1992, Elsevier Science Publishers, Amsterdam, the Netherlands

KUPITZ,JNuclear's potential role in desalinationNuclear Engineering International, 1992, December, p 41-42.

GUREYEVA,LAetalSmall power nuclear energy sources. Conversion potential.Status and perspectives.Proceedings ENS '94, Lyon, Oct 2-6, 1994, Vol 2, p 51-57.

ww victor/F-dnhpp ea


1995-02-13

Fl

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Schedule

Description

CAREM-25 (Conjunto Argentina de Ractores Modulares)

NHPP

PWR

Small power/100 MWt

ESTVAP, Argentina

Studies

The CAREM-25 reactor plant applications considered include powergeneration, industrial steam production, water desalination, and urbanheating for small grids or isolated location. The CAREM-25 reactor plantmay generate up to 25 MWe. To overcome the increased specific cost due tothe very small scale, standard design and modulized factory construction areenvisaged.

The main design features of the CAREM-25 plant are:

o Integrated arrangements of the RCS widi a once-through,helical tube steam generator which is located at the top of thedowncomer

o Natural circulation of the coolant in the primary circuit

o Steam-water pressurizer on the top of the RPV

o Shut-down condenser, in the upper part of the RPV, providesfor the removal of decay heat by the evaporator system in theenvironment for 72 hours after shut-down

o Emergency core cooling injection envelops only the upper partof the RPV with the penetration

During the LOCA event the steam generated from the core will pass into thesteel primary containment vessel and condense. Passive air cooling willdissipate the heat and condense the steam.

The basic conceptual study of the CAREM-25 plant is already completed. In1996, the CAREM-25 reactor plant will be ready for construction. Theconstruction of the 100 MWt CAREM-25 plant may start as early as 1996.

ww viclor/Fl ea

STUDSV1K ECO & SAFETY AB STUDSVnCES-95/10

1995-02-13

Fl:2(3)

CAREM-25 Unit Data


Core

25MWe

Average core power densityAverage burn-upInitial enrichmentActive core heightCore diameterNumber of fuel assembliesNumber of fuel rods per assemblyNumber of control assemblies

Reactor cooling systemPressureCore temperature, outlet/inlet

Reactor pressure vesselInside diameterHeight

Steam generatorNumberTypeFeed water pressureFeed water temperatureSteam pressureSteam temperature

ContainmentTypeDimension (height/diameter)

55kW/l22MWd/kg3.4%1.4 m1.29 m6110825

122.5 bar326/284 °C

2.84 ml l m

12once-through50 bar200 °C47 bar290 °C

Pressure suppression15.8/11.5 m

References FORSBERG, C W and REICK, W JWorld-wide Advanced Nuclear Power Reactors with Passive and InherentSafety. What, Why, How and Who.ORNL-TM-11907, USA, 1991.

HARRIAGUE, S et alThe case for very small power plants: CAREM-25.Proceedings ENC '94, Oct 2-6, Lyon, 1994, Vol 2, p 477-480.

ww victor/Fl ea


1995-02-13

Fl:3(3)

n1f1 =a'14

sI1

Mi\\ 'PI

r i!i

i

o oci0 OO

op c0£)O

=o oc

0 OO1

0 OC30 O O

0 Ou0 O < 1

0 OC0 OO

0 0 <ft f \ ( '

J -f 1 J

10

IPI

§5ai

: o

: o

0

: o

1ifRHI

Ri 'Iff- 0 l i t

>c(51Ii1

12

i ill

91

r1

Figure FlCAREM-25: Reactor coolant system

123456789

10

Hydraulic control driveControl drive rod structureWater levelC team generatorWater inletStearr. outletAbsorbing elementFuel elementCoreCore support structure

ww victor/Fl ea


1995-02-13

F2:i(3)

F2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Schedule

Description

PAES-100

NHPP

PWR

Medium power/160 -170 MWt

OKBM.RF

Commercial

The PAES-100 floating nuclear plants can supply with power the costalindustrial and remote areas and regions. Single unit concept: reactor - turbine- generator - power line are adopted for the PAES-100 power system. Theplant design allows the customer to co-generate electric power, heat anddrinking water according to the desire in any proportion. In the PAES-100power system it is supposed to use two KLT-40 nuclear steam supplysy3tems. KLT-40 (see G2) has development and operation experience ofequipments and systems in Russian ice-breakers ("Arktika", "Sibir","Russia").

Main components of the PAES-100 reactor plant are the following: reactor,fovr vertical steam generators and four primary circulating pumps areconnected by short load-bearing-nozzles and form a steam generating unit

Core fuel assemblies contain pin-fuel elements. The steam generator is a one-through heat-exchanger, the tube system of which is made of the cylindricalhelical corrosion resistant coil. The main circulation pump is a centrifugal,single stage, glandless unit with a canned two-speed motor.

Two PAES-100 floating nuclear plants are intended for the remote north andfar-east regions of Russia. Taking into account the availabilities of KLT-40NSSS with equipment and systems already developed, the work on buildingof a floating PAES-100 reactor plant may be started immediately.

ww victor/F2


1995-02-13

F2:2(3)

PAES-100 Unit Data

Main ship dimensionsLengthWidthDraughtDisplacement

Reactor plantElectric power (brutto)Reactor heat ratingReactor numberSteam temperatureSteam pressureLifetime of reactor vessel, maincomponent vessels and pipelinesStaffPrimary coolant pressure

120 m30 m6m16 000 - 20 0001

100 MWe160 -170 MWt2290 °C40 kgs/cm2

40 years60128 bar

Primary coolant temperature inlet/outlet 278/312 °C

References POLUNICHEV, V I et alNuclear Steam Supply System KLT-40 enhanced safety as independentpower supply source. Employment prospects.Proceedings Intemational Conference Nuclear Society International,27 June - 2 July, 1993, Nizhni Novgorod, RF.

MITENKOV, F MPerspectives of ship NSSS's.Atomnaja energia, April, 1994, 76, N 4, p 318-326.

VEXLER, L and PANOV, YRussians float their APWS-40 idea.Nuclear Eng Int, V 39, No. 485, December 1994, p 40-41.

ww victor/F2

STU'DSVIK ECO & SAFETY AB STUDS VIK/ES-y.V K) F2:3(3)

1995-02-13

r*T"><

1111111

/ 1

c-o-

I

Figure F2PAES-100: Layout.

12345

6

ReactorReactor coolant pumpSteam generatorTurbo generatorCondenserSecondary coolant pump

ww viclor/F2


1995-02-13

F3:i(3)

F3

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

TRIGA Power System (TPS)

NHPP

PWR

Small power/64 MWt

General Atomic. US

Commercial

The TRIGA Power System (TPS) is based on the proven TRIGA researchreactor developed by GA technologies, which has been successfully operatedduring more than 800 reactor years world-wide. The concept uses a reactorpressure vessel to achieve temperatures adequate for a district-heating net-work. The primary system consists of a reactor unit and a heat-exchangerunit A primary coolant pump is located between these two systems. TheRPV is located at the bottom of a large tank filled with water, which pro-vides an emergency heat sink. Particular safety features of the TRIGAconcept arc, on the one hand, the uranium-zirconium hydride fuel with itsprompt negative temperature coefficient, and on the other hand the transitionfrom normal operation to decay heat removal after failure of the forcedconvection without any active intervention by means of a pressure-balancedventuri nozzle.

An important feature of the TPS module design, which is motivated bysafety, is the segregation of hot water from cold water. The intent is tominimize the amount of primary coolant that is available, the flash to steam inthe event of a reactor pressure boundary rupture and thus ensure an adequatecoolant inventory for decay heat removal for accidents. The amount of hotwater involved in normal operation heat transfer is only about 9 % of thetotal reactor-free volume. The remaining water is maintained at about 71 °Cby the auxiliary cooling system, which operates continuously by naturalconvection.

Optimum TPS electric power output is achieved with a sealed supercriticalorganic Rankine cycle with a single turbine generator.

Although several arrangements for extracting heat are possible, the preferredsystem arrangement is a cogeneration heat-exchanger located in series with aprimary heat-exchanger.

ww victor/F3 ea

STUDSVIK ECO & SAFETY AB STUDSVK^S-95/10

1995-02-13

F3:2(3)

Schedule No TPS reactor plant for decentralized co-generation has been ordered yet.

Triga Power System Unit Data

Nuclear heat sourceReactor powerPrimary pressureCore inlet temperatureCore outlet temperaturePool temperaturePrimary coolant flow

Power conversion systemWorking fluidPeak net power outputTurbine inlet temperatureTurbine inlet pressureFlow rateMinimum loadCold start timeStep load changeRate of continuous load change

Desalination systemTypePeak water productionSpecific power consumption

Heat systemPeak thermal powerPeak temperature

64MWt31.2 bar182 °C216 °C71 °C418 kg/s

R-11411.8 MWe204 °C41.4 bar402 kg/s1.0 MWe+1.5 hours±15%+10%

Reverse Osmosis53 000 m3/day5.3 Wh/m3

62MWt188 °C

References SCHLEICHER, R WTRIGA power system for power, water and heat in remote communities.Potential Small Nuclear Reactors for Future Clean and Safe Energy Sources.Ed. by H Sekimoto, 1992, Elsevier Scientific Publishers, Amsterdamp 157-164.

ww victor/F3 ea

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 F3:3(3)

TO CHEMCONTROL& MAKEUPSYSTEM

CONFINEMENT

— LINER

HIGHPOINTVENT-

TO AUXILIARY M U E f VALVESERVICE TANK r , " O F 2 ! TO CAS

^ _ _ J A — » REMOVAL

PRESSURIZETM M0DULt „

HIGH POINTVENT

• * - FROM RECUPERATOR

— TO TURBINE

PRIMARYHEAT EXCHANGE

SEISMIC STOPS

Figure F3TPS: Nuclear heat source arrangement.

ww vicior/F3 ea

STUDSVTK ECO & SAFETY AB STUDSVIK/ES-95/10

1995-02-13

F4:i(3)

F4

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

ABV

NHPP

PWR

Small power/60 MWt

OKBM.RF

Detailed design

A 60 MWt floating N

Schedule

primary coolant natural convection and passive safety systems is beingdeveloped in Russia for electric power and heat production in Northern andNorth-Eastem regions.

The ABV reactor is provided with a gas pressurizing system. The steamgenerator tube system is located in the upper part of the annulus between thereplaceable unit and reactor vessel. The tube system comprises somesections, each of them comprising some assemblies. The possibility ofdisconnection is provided in the event of leaks.

Heat for prolonged emergency cooling is transferred, due to natural con-vection, to cooling system loops with pure water store evaporation (in steamgenerators) into the environment. Cooling system efficiency and waterstorage is such that residual heat removal may be carried out during severalweeks without power and water supply from outside. Complete sluggishnessof the ABV safety system operating without power and wacer consumption,is achieved due to the application of self-actuated equipment withoutconventional automatic protection circuits.

The possibility of simultaneous electric power and heat supply to users indifferent proportions along with preservation of reactor thermal outputprovides for acceptable economic efficiency for energy sources, in spite ofthe low unit power of ABV plants.

Ready for construction. One ABV reactor plant is intended for thedecentralized north region of Russia.

ww victor/F4 ea


1995-02-13

F4:2(3)

ABV Unit Data

Reactor thermal output

Vessel height

Vessel diameter

Reactor volume

Primary pressure

Temperature at core inlet/outlet

Seismic stability by

MSK-64 scale

Lifetime


Number of fuel elements in assembly


Maximum linear heat rating

Refuelling period

60MWt

4.8 m

2.6 m

not less than 8

15.7 bar

233/340 °C

9 magnitude

50 years

55

102

6

180 W/cm

3 years

References SAMOYLOV, OAdvanced Passive Pressurised Water Reactors of Small and MediumCapacity - Russian Approach.Proceedings 9th Pacific Basin Nuclear Conference, Sydney, 1-6 May, 1994.

ww victor/F4 ea

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/ll) F4:3(3)

1995-02-13

Figure r4ABV: Power

123

456789

unit schematic.

ReactorControl rod driveAccumulator of emergencycooling down systemsHeat exchangerPurification system pumpTurbineIon-exchange filterPurification system coolerPressurizer

1011121314151617

BubblerBubbler coolerFilterVentilation tubeProtective shellWater store tankFlushing-through system pumpLoop heat exchanger

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1995-02-13

F5:i(3)

F5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

ELENA

NHPP

PWR

Very small power/3 MWt

RRC-KI, RF

Basic design

The unattended self-contrintended for the supply of a small settlement with heat (up to 3 MWt) andelectricity (up to 0.1 MWe). This reactor can also be supplied with adesalination facility. The DNHPP "Elena" does not require humanintervention during its total service lifetime (about 25 years). Its techno-logical scheme does not include valves, mechanisms and devices requiringany operations and maintenance during lifetime. During its lifetime the plantoperates on the initial core loading.

The control and safety system including control rods, drive mechanisms,sensors etc only serves for the start-up and the shutdown of the reactor aswell as for the trips under emergency conditions. The main features of theplant are the following: low power, small size, low power density,underground location.

A reactor core consisting of the standard VVER-type fuel elements. A levelof the negative temperature reactivity feedback makes it possible tocompensate for the burnup effects by an insignificant reduction of the coolanttemperature. Therefore the reactor does not need the excess reactivity for theburnup. There is natural convection of the primary and secondary coolantDue to a low core power density (~3kW/l) the reactor does not requireemergency cooling. Use of the thermoelectric energy conversion permits thereactor plant to be substantially or significantly simplified.

The reactor and thermoelectric generators are accommodated inside the pro-tection vessel filled with water of the secondary circuit. Due to the use ofthree strong steel vessels, containment and air-tight concrete box, thepossibility that the nuclides will penetrate into the environment is ruled outAlso, the "Elena" type power source for deep sea application is underconsideration in RRC-KI.

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1995-02-13

F5:2(3)

Schedule Ready for construction. No "Elena" reactor plant has been ordered yet.

ELENA Unit Data

Power supplied to a district heating system

Electric power

Temperature of water supplied to a districtheating system

Power unit size, diameter/height

Weight (without primary and secondary coolants)

Distillate output

Distillate salt content

Operational life

up to 3 MWt

up to 100 kWe

up to 90 °C

4.5/12 - 14 m

1601

up to 60 m3/h

less than 25 mg/1

up to 25 years

References Thinking smallNuclear Engineering International, 1992, November, p 11-12.

KAPLAR.EetalUnattended self-controlled nuclear thermoelectric low power plants fordecentralized district heatingMoscow, RF, 1994, Report RRC "Kurchatov Institute".

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STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/K F5:3(3)

1 Core2 Protection screens3 Strong vessel4 Additional ensurance shell guard vessel5 Protection vessel6 Containment8 II-III circuit heat exchanger9 Heat exchanger of district heating system

10 Heat exchanger for residual heat removal11 Pressurizer (or residual hat removal system)12 Valves13 Hot water supply pipes14 Prime circuit pressurizer15 Second circuit pressurizer16 Concrete box wall17 Concrete box top18 Auxiliary units room19 Control system driver20 Electric power output

Figure F5ELENA: Schematic.

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1995-02-13

F6:i(3)

F6

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

Compact HTGR Gas Turbine

NHPP

PWR

Small power/29 MWt

General Atomic, US

Studies

The gas turbine has the highe

Schedule

available technology, but it requires gas inlet temperatures that can only bedelivered by an HTGR core with TRISO-type particle fuel.

The main elements of the Compact HTGR Gas Turbine system for remotesites are the reactor core, turbo compressor, power turbine, recuperator,precooler and generator. This design has an integrated arrangement Thecore is surrounded by a combined graphite BeO reflector material. Controlrods are located in the reflector.

In the 10 MWe study the turbo machinery was conveniently divided into twohorizontal sections, a free-running turbo compressor, the function of which isto provide the energy to transport the helium through the primary system anda separate power turbine driving the generator. A very compact 1 MWe unitis best suited to a single shaft vertical unit

No Compact HTGR Gas Turbine plant has been ordered yet This programhas a low funding level. As any gas turbine, the HTGR concept will haveunique safety problems resulting form turbine overspeed or shaft failure andrequires a decision on whether or not an intermediate heat-exchanger isneeded for an early demonstration project.

ww victor/F6 ea

STUDSVnC ECO & SAFETY AB

Compact HTGR Gas Turbine

Electric output

Core thermal power


Core diameter

Core height

Core power density

Fuel element type

Number of elements

Approximate core life

Turbine pressure ratio

Turbine speed

Compressor speed

Vessel diameter

Vessel height

STUDSVIK/ES-95/10

1995-02-13

rbine Unit Data

lOMWe

29MWt

850 °C

1.4 m

1.4 m

13.5 w/cm3

Cylindrical

175

7 years

2.0

8 000rpm

36 000 rpm

3.0 m

7.0 m

lMWe

3.3 MWt

850 °C

0.67 m

0.67 m

14.0 w/cm3

Cylindrical

42

7 years

2.0

12 000 rpm

12 000 rpm

1.5 m

6.1m

F6:2(3)

References SCHLEICHER, R W and WISTROM, J DSmall reactors utilizing high temperature capability of coated particle fuel.Proceeding International Meeting on Potential of Small Nuclear Reactor forFuture Clean and Safe Energy Sources, Tokyo, Oct 23-25,1991.

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STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 F6:3(3)

1995-02-13

COMPRESSORINLET

TURBINEINLET

EMERGENCYCOOLING COILS-

PRECOOLERECONOMIZER

FUEL ELEMENT

SIDE SHIELD

CORE BARREL-

REFLECTOR

\ \ COMPRESSOR OUTLET

TORE GUPPOhT GRIO

BOTTOM SHIELD

CONTROL ORIVE

Figure F6Compact HTGR Gas Turbine: Reactor vessel and internals.

ww victor/F6 ea


1995-02-13

G

GO

Ship Reactors

Introduction

Civil nuclear ships have various excellent characteristics compared to con-ventional ships. The use of nuclear energy opened the possibility of designingships of a high ice-cutting capability due to the absence of power restrictionsand practically unlimited operation time (typically 400 days with a powerutilization factor of about 0.6). Experience has shown that these ships areeconomically competitive with diesel ice-breakers. It gives a strongmotivation to develop marine reactors and nuclear ships. At present, twomain applications are relevant for ship mobile systems (SMS):

oo

For ships navigating on sea surfaceFor power source used in deep sea

These concepts are now under development in Russia and Japan.

Russia has had excellent experience for about 30 years in the field of ice-breakers. At present, Russia operates ice-breakers of the "Arctica"-type.Their capacity is from 34 to 49 MW. The Russian concept of Arcticnavigation development includes the necessity of the following classes of ice-breakers:

o 30-35 MW ice-breakers for shipping in shallow waters of the

Arctic coast and Siberian rivers

o 60 MW ice-breakers for piloting transport ships

o 110 MW ice-breakers designed for reliable all year round shippiloting

In Japan, after successful demonstrations of the operation of the nuclear ship"MUTSU", new research and development programs are now under wayconcerning nuclear ships and mobile systems. In JAERI the basic concepts oftwo types of marine reactors have been developed: the first one is for an icebreaking scientific observation ship and the latter is for deep-sea research.

References

MITENKOV, F MPerspectives of ship motor plants.Atomnaja Energia, 1994, v 76, N4, April, p 318-326.

SEKIMOTO, H (Editor)Potential of small nuclear reactors for future clean and safe energy sources.1992, Elsevier Publishers, House, Amsterdam.

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STUDSVK ECO & SAFETY AB STUDS VK/ES-95/10

1995-02-13

G l

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

OK-900A

SMS

PWR

Small power/170 MWt

OKBM.RF

Commercial

Description

Schedule

The first generation of naval reactor plants was represented by an OK-150loop type PWR that operated on the "Lenin" ice breaker in 1959-1966 andwas subsequently replaced by a more modern one of OK-900 type NSSS.

Modular reactor plants of the OK-900 type have been developed for the nextgeneration of ice breakers of the "Arctica" type. These systems are radicallydifferent from those of the first generation in their technical operational andeconomic parameters. Main components of the OK-900A NSSS includingsteam generators and primary circulating pumps are connected by short loadbearing-nozzles and form a steam generating unit

The reactor has inherent safety properties due to:

o Negative reactivity coefficients of the core

o Passive principle of reactor shutdown by control clusters

o High heat accumulating capacity

o Elimination of large diameter pipelines in the primary circuitand insertion of devices limiting coolant discharge at pipelineruptures.

Long tenn operation of OK-900A NSSS in ice breakers has demonstratedthat they completely satisfy the necessary requirements. The service time ofNSSS equipment on the "Arctica" and "Siberia" ice breakers that have beenin operation since 1974 and 1978, respectively, is close to 100,000 hrwithout any replacements of the system equipment. There have been noreactor plant accidents that have resulted in even a temporary interruption ofthe ice breaker operation.

No new OK-900A reactor plant has been ordered yet.

ww victor/Gl ea


1995-02-13

Gl:2(3)

OK-900A Unit Data

Type-

Number or reactors

Independence with one fuel loading

Total shaft power

Rated thermal reactor power

Primary pressure

Primary coolant temperature,inlet/outlet

Steam production

Secondary system- temperature- steam pressure

Ice breaker

OK-900

2

1 050 days

44 000hp

2x159 MWt

128 bar

277/320 °C

2x220 t/h

305 °C30.9 bar

"Lenin"

OK-900A

2

1 050 days

75 000 hp

2x171 MWt

128 bar

277/320 °C

2x240 t/h

290 °C32.4 bar

"Arctica","Siberia","Russia","Sov.Union","Oct.Revolution"

References VASYUKOV.VIetalExperience in development and operation of NSSS for civil ships;Issues of life extension.Proceedings TANS, 1992, St Petersburg, RF, p 47.

ww vicior/Gl ea

STUDSVIK ECO &. SAFETY AB STUDSVIK/ES-9.V10 Gl:3<3>

1995-02-13

Figure ( i lOK-900A: Reactor coolant svstem.

1-1

345

7

Expansion (ankLow pressure air systemDrainageAir removal systemMakeup and emergency spray systemPressure systemSafety device

«9

if)1112

n

Water inlet, outletPressurizer syslemWater inlet, outletWater inlet, outletMCPIHX

ww victor/Gl ea


1995-02-13

G2:i(3)

G2

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

KLT-40

SMS

PWR

Small power/170 MWt

OKBM.RF

Commercial

Description The new generation of naval reactor plants represented by KLT-40 andKLT-40M reactor plants, designed for a new series of ice breakers of the"Timyr" type and "Sevmorput" large carrier corresponding to moderninternational requirements of nuclear safety KLT-40 reactf"" plants, wasdeveloped with the maximum use of the proven components of OK-900NSSS(seeGl).

The main differences between the KLT-40 and OK-900 reactor plants dealwith the improvements related to the higher safety of nuclear facilities, theirmaintainability and better operating lifetime. In particular, the reactor designis simplified, measures are taken to improve the monitoring of the reactorvessel, its cover, drives of the control and safety system. The mainmodernization measures were implemented in the reactor plant structurecovering cleaning system ducts, cooling down system and the pressurizer,and measures to reduce emergency coolant leakages.

Main safety systems of KLT-40 ar**-:

o Control and protection system (three redundant supply lines)

o Liquid absorber injection system

Residual heat removal system (through steam generator orthrough the heat-exchanger into the sea or air - twoindependent channels)

Emergency core cooling (two independent channels:hydroaccumulators and electric pumps for water injection)System of emergency pressure reduction in the containment

Containment system.

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1995-02-13

G2:2(3)

Schedule

A more powerful reactor of the KLT type - KLT-3 reactor plant withthermal output 300 MWt is also developed by OKBM.

No new KLT-3 or KLT-40 plants have been ordered yeL

KLT-40 Unit Data

Type

Thermal power

Electric power

Primary pressure

Primary coolant temperatureinletoutlet

Steam production

Steam temperature

Steam pressure

Feed water temperature

KLT-40

170 MWt

50MWe

128 bar

277 °C320 °C

240 t/h

305 °C

31 bar

104- 108 °C

KLT-3

300 MWt

90MWe

146 bar

274 °C322 °C

400 t/h

295 °C

40 bar

194 °C

References PULUNICHEV.VIetalNuclear Steam Supply System KLT-40 enhanced safety as IndependentPower Supply Source. Employment prospects.Proceedings TANS, 1992, St Petersburg, RF, p 173.

ww victor/G2 ea

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 G2:3i3)

1995-02-13

Figure (i2KLT-40: Reactor plant.

1 Reactor2 Steam generator3 Heal exchanger4 Metal/water shielding syslem

lank5 Pressure suppression pool6 Biological shielding7 Protective enclosure walls8.9 Valves

101112

n14151617IX

Air cooling heal exchangersCondensate pumpsSea waierCondensersWater storage tanksEmergency teed waler pumpsHydroaccumulalorsECCS pumpsThird circuit

ww victor/G2 ea


1995-02-13

G3:i(5)

G3

Format

Description

Title


Reactor type

Power output

Organization (name)

Development status

Description

MRX

SMS

PWR

Small power I

JAERI, Japan

Basic design

JAERI has coJAERI has conducted a design study of the Marine Reactor X (MRX) forice-breakers. It has a 60 000 horse power propulsion system with two 100MWt MRX reactor plants.

MRX is an integral type PWR and employs in-vessel type control rod drivemechanism and water-filled containment vessel including their reactor vesselsurrounded by thermal insulation. At the upper part of the containmentvessel, heat pipe type coolers are placed. Dip plates are also placed in orderto enhance steam condensation in the event of loss of coolant accidents formitigating containment water slashing.

The core consists of 19 fuel assemblies. Thirteen assemblies contain controlrod clusters. Six are for reactivity control during operation and the otherseven are for back-up reactor shutdown. Standard fuel pins with Zircaloycladding (9.5 mm outer diameter), and burnable poison for reducing the peakfactor, are also applied. Hexagonal arrangements are preferred rather thanrectangular ones.

There are only eight pipes (50 mm diameter) carrying primary coolant andpenetrating the reactor vessel. They include four for RCS/RHRS, two forpressurizer safety valves, one for aressurizer relief valve and one forpressurizer spray.

Steam generator tube inspection can be carried out from the secondary sideafter opening blind flanges on steam and feedwater headers. Main coolantpumps are inspected in an inspection pit After the pumps are withdrawnhorizontally from the vessel they are transferred to the pit

ww victor/G3 ea


1995-02-13

G3:2(5)

Schedule

Only small LOCAs can be possible in the MRX. In such a LOCA, theengineered safety system keeps the core flooded and removes decay heatwithout emergency water injection.

Maintenance procedures for main components arc now being studied. NoMRX reactor plant has been ordered yet

MRX Unit Data

Reactor power

Reactor type

Reactor coolantOperating pressureInlet/outlet temperatureFlow rate

Core/FuelEquivalent diameterEffective heightAverage linear heat rateAverage power densityFuel typeU-235 enrichmentFuel inventoryFuel life timeFuel average bumupNumber of fuel assembliesFuel rod outer diameter

Control rod drive mechanismTypeNumber of CRDMs

Main coolant pumpType

Number of pumps

Steam generatorType

Tube materialTube outer diameterTube inner diameterSteam temperature/pressure

100 MWt x 2 units

Integral type PWR

120 bar282.5/297.5 °C4 500t/h

1.45 m1.40 m74 kW/m43kW/lUO2 fuel rod4%6.5 t8 years25 GWd/t199.5 mm

In-vessel type13

Horizontal axial flowcanned motor type

2

Once-through helicalcoil typeIncoloy 80019 mm14.8 mm289 °C/40 bar

ww victor/G3 ea


1995-02-13

G3:3(5)

Steam flow rateHeat transfer area

Reactor pressure vesselInner diameter/height

168 t/h970 m2

3.7/9.3 m

Containment vesselInner diameter/heightDesign pressure

Overall length

Water line length

Maximum breadth

Water line breadth

Moulded depth

Full load draught

Full load disp

Continuous ice breaking ability

Ship crew

Observation personnel etc

Propulsion shaft horsepower

6.8/13.0 m40 bar

150 m

141m

30 m

28 m

15 m

9m

22 0001

2.1 m thickness forlevel ice

60 persons

40 persons

20 000 ps x 3 shafts

References SAKO.KetalAdvanced Marine Reactor MRX.Proceedings International Conference Design and Safety of AdvancedNuclear Power Plants, Oct 25-29, 1992, Tokyo, Vol 1, p 6.5.

ww victor/G3 ea

STUDSVIK ECO & SAFETY AB STUDSVlK/TiS-95/K) G3:4(5)

Figure (i3MRX:

ww


1995-02-13

G3:5(5)

Containment vessel(Inner D: 6.8mInner H:13.0m)

Containment ••tercooler (4 trains)(Heat pipe type)

Dip plate(for promotesteam condensing Swater stabi Iizing)

Emergency decayheat removalsystem (x4)

Plate for waterstabi Iizing(x8)

(N} gas)

Rated water level

Main steam Iine(2 trains)

Shielding

Water-tightshe I I

Thermal insulator(Mirror insulationtype)

flelding

Control rod drivemechanism(x13)

later sprayheder

Pressurizerheater

Main coolantpump(x2)

Pressure relvalve (x3)

ief

~ Steam generator(Once-throughhel i ca I coi Itype:2 trains)

— Reactor vessel(Inner D:3. 7 m )

— CoreFuel ass. (x19)

Flow screen

ww viclor/G3 ea

STUDSVJX ECO & SAFETY AB STUDSVK/ES-95/10

1995-02-13

G4:i(3)

G4

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

DRX

SMS

PWR

Very small power/0.75 MWt

JAERI, Japan

Basic design

The Deep-sea Reactor X (DF

Schedule

150 kWe for undersea power sources, which can provide enough powercapacity for a bathyscaph. It is a PWR with an integrated arrangement with asteam generator and turbine inside the reactor vessel. The combination oftwo spherical shells 2.2 m in diameter, made of titanium alloy, basicallycomposes a pressure shell.

The core can be operated for more than 400 days at 750 kWt, whichcorresponds to about 4 years operation without changing of the fuelassuming 30 % of effective load factor. The core consists of a single fuelassembly with a 36.8 cm equivalent diameter and a 34.4 cm effective heightThe U-235 concentration of about 11 % is necessary to maintain criticalityduring 5 500 MWd/t burn-up.

The BWR type DRX is also under consideration in JAERI.

No MRX reactor plant has been ordered yet. JAERI has been conductingdesign studies of the DRX reactor plant.

ww victor/G4 ea


1995-02-13

G4:2(3)

DRX Unit Data

Maximum operating depth

SpeedAdvanceVertical, transverse

Overall length

Breadth

Moulded depth

Continuous operating period

Reactor output

Generator output

Pressure shell inner diameterReactor containerLiving room

Number of crews (incl. scientists)

6 500 m

3.5 kn1.5 kn

24.5 m

4.5 m

6.0 m

30 days

750 kWt

150 kWe

2.22.8 m

References IIDA, H et alDesign study of deep sea reactor DRX.Proceedings International Conference on Design and Safety of AdvancedNuclear Power Plants, Oct 25-29, 1992, Tokyo, Vol 1, p 6.6.

ww victor/G4 ea

[>-••-•• v •.

A

\ ' '

• : : • ' . , ' ( . -

' ' • <

rVt---i7:1 !

a-~ -' r ':

i-1 \


1995-02-13

1:1(4)

I Space Nuclear Reactors and Isotope Batteries

10 Introduction

The field of space reactor designs is one of the most innovative areas innuclear engineering and technology.

Since the start of the space age in the 50's, a set of nuclear power supplyoptions has been developed by the US and Russia to support the military andcivilian space programs.

The investigation and development of space nuclear reactors has mainlyfollowed two lines:

oo

Power sources for electricity generationNuclear propulsion

Space nuclear plants, which produce complex propulsion and electricalpower are also discussed.

A generic classification of space nuclear power source types is depictedbelow:

Nuclear power systemtype

RadioisotopeThermoelectricGenerator (RTG)

Radioisotope dynamicconversion generator

Reactor systemsHeat pipeSolid coreThermionics

Reader systemHeat pipeSolid core

ReactorSolid corePellet bedFluidized bedGaseous core

ww victor/1-snr ea

Electric power range(module size)

up to 0.5 kWe

0.5kWe-10kWe

10kWe-lMWe

1 - 10 MWe

10- 100 MWe

Power conversion

Static: Thermoelectric

Dynamic: Brayton ororganic Rankine cycles

Static: Thermoelectric,thermionicsDynamics: Brayton,Rankine or Stirling

Brayton cycleRankine cycleStirling cycle

Brayton cycle(open loop)

StirlingMHD


1995-02-13

1:2(4)

Here, the primary system output is electrical energy, which is produced byconverting radioisotope decay heat or the thermal energy released in thenuclear fission reactor. For conversion to electric power the static(thermoelectric or thermionics) or dynamic (e.g. Rankine or Brayton cycles)principles are used. For space power applications in the MWe and beyondrange there is a number of interesting innovative nuclear reactor technologyoptions.

From the late 1950's substantial efforts in different countries (primarily theUS Rover-NERVA program) have also devoted to nuclear propulsion.

Nuclear rocket propulsion is attractive because nuclear heat generated in ahigh power density reactor core can be used to heat a low molecular weightpropellam to high temperature, which enables a much higher propellantexpulsion velocity than can be obtained in a chemical engine.

Three different categories of innovative reactor concepts have been con-sidered for nuclear propulsion (solid core, liquid core and gas core). Eachcategory contains several design concepts. In theory the gas core wouldproduce the highest temperature, the liquid core intermediate temperatureand the solid core lower temperature.

In 1973, the Rover-NERVA programs (based on solid core thermal engine)were terminated as a consequence of an indefinite delay of plans for spaceinissions which would call on the capability of nuclear propulsion.

Since 1961, the US has launched another twenty civil and military nuclearspace power systems. Russia had flying nuclear power systems in space sinceabout 1965. These nuclear power sources have very small power nuclearreactors using thermoelectric and thermionic conversion principles (e.g. USSNAP-type and Russian Romashka and TOPAZ-I.II-types).

Russia has during the period 1970 to 1988 launched 31 space-crafts withthermo-electric nuclear reactors. In the late 80's Russia launched two space-crafts with the thermo-emission nuclear reactor Topaz 1. Later theydeveloped the Topaz II thermo-emission nuclear reactor which has not yetfound its application, but is currently tested in cooperation with the USA.

The USA have mainly used plutonium-238 batteries and have only launchedthe SNAP 2 nuclear reactor. The development of the US SP-100 nuclearreactor program, which started ten years ago, has recently been terminatedand the technology achievements are nor transferred to the industry.

Some other countries (e.g. France and Japan) are also investigating thepossible development of space nuclear reactors..

Further space activity will require even larger quantities of safe, reliable andautonomous power sources.

ww victor/I-snr ea

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 1:3(4)

1995-02-13

The following key issues can be addressed in the research field:

o The physics of thermionic conversion (surface and near-surfacephenomena on electrodes, the effect of modification of thesurface on the operation of the output, plasma processes in theinterelectrode gap, the effects on the thermionic energy con-version with fission products in the interelectrode gap

o The research and development of fuel composites (includingnitride and carbonitride fields, processes of fuel-cladding inter-action, swelling processes, wastage, recondensation, and fissionproduct behaviour)

o Improvement of the structure (including geometrical optimi-zation, selection of electrode materials, fission products release,methods of supplying Cs to the interelectrode gap, providingelectrical insulation, switching, necessary thermal modes, andso on)

o Reactor and structural materials (moderator, reflector,absorbers, thermal and electro-insulation, and cement com-pounds)

o Coolant technology (corrosive processes and mass transfer inloops, methods and means of controlling the action ofadmixtures, and diagnostics and the prevention of leaks)

o Improvement of methods for releasing waste heat (heat pipes,possible designs of radiators, liquid droplet, and other newtypes of radiators)

o Measures to ensure lifetime

o Diagnostics, identification, and information during ground andin-flight testing (sensors, instruments, methodology, andalgorithms)

o Guaranteeing nuclear and radiation safety

For the US and Russia 1994 -1995 governmental proposals cut the researchand development funding support and related facility funding for nuclearreactors that have no commercial or other identified application. At presentthe USA will perform the Cassini program for investigation of Satumus andwill use plutonium batteries as the internal power source.

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STUDSVJX ECO & SAFETY AB STUDSVK/ES-95/10

1995-02-13

1:4(4)

References

o

ANGELO, J A and BUDEN, DSpace nuclear power.Orbit book company, inc. Malabar, Florida, US, 198S.

BOHL RJetalNuclear rocketry review.Fusion Technology, 1991, December, v 20, N4, part 2,p 698-710.

MARSHALL,ACStatus of the nuclear safety assessment for the TOPAZ-H spacereactor program.Proceedings ICENES '93, Makuhari, Chiba, Japan, 20-24 Sept,1993, p 351-358.

PONOMAREV-STEPNOI, N NNuclear energy in space.Space Nuclear Power Systems, 1989.Orbit book company, Fl, 1992, v 11, p 437-440.

CARRE, FetalSpace nuclear power system studies in France and envisagedmissions.Proceedings Nuclear Power Engineering in Space, Obninsk,RF, 15-19 May, 1990, p 136.

MILOV, Y GNuclear Power in the Russian Space Program.AIP Conference Proceedings No. 32412th Symposium on space nuclear power and propulsion.1995, January 8-12, USA, NM, Albuquerque.

Thermionic Energy Conversion Conference Proceedings, 1993,May 5-7, Sweden, Göteborg.

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1995-02-13

I I : K2)

II

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

GPHS-RTG (General Purpose Heat S(

SNR, power source

Radioisotope thermoelectric generator

Very small power/0.3 kWe

Westinghousc, US

Commerical

The world's first "atomic battery" was

Schedule

Space Nuclear Auxiliary Power (SNAP) program. The RTG utilized thenatural decay heat of a radioactive isotope, converting the heat directly intoelectricity via thermoelectrics.

The advanced GPHS-RTG design is a general-purpose heat source andconsists of a graphite block with Pu-238 pellets. Each source has a poweroutput of 0."* kWe. These RTGs were flown on the last two missions,Galileo am' *, ysses. For lunar outpost applications interchangeableradioisof jt at source assemblies, the thermal source that drives theenergy c /. rr £i, may be developed. This concept allows heat sourceassemb'.; m oe transferred between various generator units, therebymaximi J. y tilization of the Pu-238 supply.

Advance oncepts, utilizing dynamic systems such as Sterling, or advancedstatic or- »ersion are required to provide the lower powered mission pro-jected ' NASA. These advanced converters will also minimize the amountof Pu ? fj required by their increased efficiency.

As Ion? is the US have programs for which "nuclear batteries" are powersource, .he DOE will be a key player in building and supplying powersource:

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GPHS-RTG Unit Data

Power output

Operation life

Weight

Output voltage

Envelope, diameter/longitude

Hot junction temperature

Fuel

Thermoelectric material

Magnetic field

On pad

Storage life

Auxiliary cooling

290 We (BOM)

2S0 We (minimum at EOM)

40 000 hours

54.1 kg

28(±0.5) V

46/< 115 cm

1275 K

83.5+1 % 238PuO2

SiGc

<I0xl0" 9 T/m

30 days unattended capability

1 year ground storage

375 K average outer shell temperature

BOM = Beginning Of Mission and EOM = End Of Mission

References COOK, B A and TANELL, W HTechnical Issues Surrounding continued Supply of RadioisotopeThermoelectric Generator for NASA Programs.Transaction American Nuclear Society, Space Nuclear Power, 1992,p 22-23.

/- OUliH SHIH AiStWBlY

I; PRiSSuRlRllluotvics

RIGWOUMISi. KANGi

M i l l I ' ' O i lINSJ1A1K». J.IC0UPU

Figure IIGPHS-RTG: Layout.

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12:1(4)

12

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Schedule

Description

ROMASHKA

SNR, power source

Fast-neutron uranium carbide reactor

Very small power/0.3 -10 kWe

RRC-Kl, RF

Detailed design

For experimental verification of the possibility of creating very small reactorswith the direct conversion of thermal power into electric power, the experi-mental reactor ROMASHKA (Camomile) was created in 1964 in RRC-KI(Russia). For ROMASHKA high temperature design the heat generated inthe core is transferred via heat conducting materials to the thermoelectricconverter with high ttmperaturc silicon-germanium elements. The reactorcore consists of UC2 fuel discs. The ROMASHKA reactor has been workingsuccessfully for 15 000 hours (-6100 kWh of electricity).

The following study of the state of the ROMASHKA reactor plant elementsshowed that the achieved parameters and lifetime were not maximum andthat they could be improved by some changes of the structure andparticularly by using flat module thermionic elements between the core andradial reflector instead of thermoelectric converter.

ROMASHKA reactor plant key features are the following:

o Maximum compactness, low mass, high efficiency and

reliability

o Complete statics of operation and absence of the coolant loops

0 Self-correction due to the negative temperature reactivitycoefficient

0 Possibility of preliminary testing.

Today the ROMASHKA reactor plant is in the focus of the ThermionicCritical Technology (TCT) Program initiated by the USA Air Force for thebimodal concept with the electric output of about 25 kW in power mode,and electric power of up to 15 kW when propuslive thrust is produced in therange from 80 to 200 N.

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ROMASHKA Unit Data

Converter-type

Power output

Fuel type

Enrichment

Reflector

Total temperature reactivity effect

lifetime

Core loading

Mass

TemperaturesCore, maximumHot junctionBase of radiator, averageDifferential across converter

Neutron spectrum

Average flux

Converter characteristicsMaterialFigure of meritMaterial efficiencyOverall efficiencyWorking voltage

Thermoelectrical/thermionic

0.3-10kWe

UC2

90 % U-235

Beryllium

-2.7%

15 000 h

49 kg

435 kg

2 173 K1253 K823 K-315 K

Fast

-9xl0 1 2 n/cm 2 / s

SiGeN/AN/A- 1 . 5 %1.6 V/section

References DONOVAN, B D and LAMP, T RA Thermionic Converters Success Story.Proceedings 10th Symposium on Space Nuclear Power and Propulsion.American Institute of Physics, New York, USA, 1993, p 1369.

PONOMAREV-STEPNOI, N N e aConceptual design of the bimodal nuclear power system based on the"Romashka" type reactor with thermionic energy conversion system.AIP Conference Proceedings No. 324.12th Symposium on Space Nuclear Power and Propulsion, 1995,January 8-12, USA, NM, Albuquerque.

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J

STUDSVIK ECO & SAFETY AB STUDSVIK/ES-9.VIU 12:3(4»

1W5-O2-13

Thermal insulator

Thermal insulator

Reflector cup(thermal insulation)

I ^ Shims

' End face reflector

Container reflector

Converter radiator

Fuel disc

Be reflector

Control rod

Vessel

End face reflector

Figure 12ROMASHKA: Reactor plant.

ww victor/12 ca

aIS)

A A

n v it

Figure 12ROMASHKA: Type biumodal reactor.

\ti A

1,

34S6

139

- Lid- End reflectors- Vessel- Rod- Fuel element-TEC

7X10.22111213

- Bushing- Electric bus- Coolant headers- Nozzle- Insert- Thermal insulation

15,211617IX1920

- Hydrogen header- Safety rod poison- Side reflector- Disc- Spring- R i n g

§V)

S1

H

a00

1/5i

sC•J\


1995-02-13

13:1(3)

13

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

TOPAZ-n

SNR, power source

Zirconium hydride moderated reactor with in-core single-cell thermionicconverter.

Very small power/5 - 40 kWc

RRC-KI, RF

Commercial

The TOPAZ-II reactor design with 4.5 - 6 kWe output is a small zirconiumhydride moderated reactor with single-cell in-core thermionic convertersdeveloped by Russia. The major subsystems that comprise the power systemare the nuclear reactor, which contains the thermionic converters, the radi-ation shield, the coolant system, the cesium system and the instrumentationand control (I&C) system. The TOPAZ-H core consists of 37 thermionic fuelelements, loaded with uranium dioxide with enrichment 96% U-23;. Fuelelements are placed into vertical holes within the cylindrical moderatorblocks. The core is surrounded by a beryllium reflector. The TOPAZ-IIreactor plant is controlled by the rotation of external beryllium drumscontaining segments or boron carbide. The basic components of the coolingsystem include: electro-magnetic pump, volume accumulator, gas absorber,bellows and radiator. The coolant is eutectic NaK. The coolant flowsthrough the radiator to the lower collector. Leaving one, it is split into twoopposite flows, which enter the shield. Then, each of the two coolant pathsbranches into three coolant pipes, enters the electro-magnetic pump throughsix coolant pipes before entering the lower plenum of the reactor. Thecoolant enters six inlets to the lower coolant plenum of the reactor beforepassing through the thermionic fuel element flow channels.

The power output and total thermal efficiency increase are realized due tothe growth of the emitter temperature. In the range of 6 - 40 kWe the curren.design can be conserved without any essential modification because its massand specific power characteristics lay in acceptable diapason.

The main features of the TOPAZ-II reactor are:

o Possibility of introducing special electrical heaters instead offuel load for testing

o Simplicity of gas fission product release

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13:2(3)

Schedule

o Possibility of refuelling at finally constructed system

The US Ballistic Missik Defence Organization (BMDO) is investigating thepossibility of launching a Russian TOPAZ-II space nuclear power system.The intended application for the TOPAZ-II reactor is the Nuclear ElectricPropulsion (NEP) Space Test Mission. The program is referred to as theNuclear Electric Propulsion Space Test Program (NEPSTP). The primarymission goal is to demonstrate and evaluate nuclear electric propulsiontechnology to establish a capability for future civilian and military missions.Also, in Russia, a conceptual development of a new space system, Topaz-Star, with the electric power 40 kW, has started.

TOPAZ-II Unit Data

Power output

Conversion type

TFE design

Coolant

Outlet temperature

Inlet temperature

Voltage of working section

Lifetime

Dimensionlengthmax diameter

Mass of reactor block

6kWe

Thermionic

Single-cell

NaK

600 °C

500 °C

28 - 30 V

>3 years

3900 mm1400 mm

1 tonne

References PONOMAREV-STEPNOI, N N et alNuclear Energy in Space.Proceedings 6th Symposium on the Space Nuclear Power Systems,Albuquerque, NM, USA, 1989.

MARSHALL, A CSums of the nuclear safety assessment for the NEPSTP (TOPAZ-II) spacereactor program.Proceedings ECENES '93, Makuhari, Chiba, Japan, 20-24 Sept, 1993,p 351-355.

MILOV, Y GNuclear power in the Russian space program.AIP Conference Proceedings No. 324,1995, January 8-12, USA, NM,Albuquerque.

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1995-02-13

13:3-•>(-•>!

130 nan

67 mm

37 mm

Volume accumulator

T

Radiator

Figure 13TOPAZ-II: Reactor layout.

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14:1(2)

14

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

STAR-C

SNR, power source

Graphite moderated uraniumconversion process

Very small power/5 - 40kWe

Rockwell International, US

Basic design

A 40 kWe Space Thermionicwith electric output 5 - 4 0 kWe has been developed in response to evolvingAir Force space power requirements in the US. The STAR-C reactor plant isbased on out-of-core thermionic power conversion elements. The STAR-Creactor core consists of hot pressed UC2 fuel plates held in graphite fueltrays. There are two core/converter configuration options. The solid core hasconverters placed adjacent to its outside surface. In the annual core con-figuration, the centre of solid graphite core is removed, providing an area forconverters to be also placed within the intericr core surface.

The out-of-core static power conversion subsystem consists of a modulararray of series and parallel-connected thermionic converters.

The separation of the nuclear heat source and the conversion system allowsfor a very flexible, rigorous, non-nuclear systems test in which the entirepower conversion subsystem can be tested for efficiency and functionalityunder operating, transient and hostile conditions. This reduces the cost andthe risk schedule of the development program.

The moving reflectors are used for reactivity control in the STAR-C powersource. Radial reflectors exist in the area between converters. The resultingneutron leakage is sufficient to produce the reactivity swing necessary forcontrol. In addition to the moving reflectors, the STAR-C design utilizes anadditional, independent shutdown system in the form of B4C rods, located inthe core.

Converter components operate in regions of low neutron fluence, whichallowed to decrease swelling below the existing ROMASHKA database.

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14:2(2)

Schedule The STAR-C reactor design is in focus of the Thermionic CriticalTechnology (TCT) program initiated by US Air Force.

STAR-C Unit Data

Thermal power

Reactor output

Net electric power

Net system effect

Peak fuel temperature

Core surface temperature

Emitter temperature

Collector temperature

Main radiator area

Mass

Lifetime

340 kWt

42.8 kWe

40.9 kWe

12.0%

2 150 K

2000K

1854 K

1031 K

5.9 m2

2 502 kg

10 years

References BEGG, L L, WUCHTE, T H and ODDING, W DSTAR-C Thermionic Space Nuclear Power System.CONF 920104,1992, American Institute of Physics.

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15

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

SP-100 (Space Power)

SNR, power source

Fast spectrum reactor

Very small power/105 kWe


Basic design

The SP-100 Space ReactorElectric under contract to the DOE to provide electric power in the range ottens to hundreds of kilowatts. The SP-100 represents an enabling technologyfor a wide variety of earth orbital and interplanetary science missions, nuclearelectric propulsion stages, and lunar/Mars surface power for the SpaceExploration Initiative.

The major elements of an SP-100 system are:

o A fast-spectrum reactor made of a niobium base alloy with aceramic form of uranium nitride fuel

o A set of actuators and controls that position the neutronreflectors and the safety rods automatically, or by command

o A shadow shield (lithium-hydride/tungsten) that attenuates bothneutrons and gammas and protects the spacecraft electronicsfrom the radiation emitted by the reactor

o A heat transport loop that transfers heat from the reactor to aset of power conversion devices by means of a liquid metal(lithium) pumped by a self-actuated electromagnetic pump, anda heat rejection loop that transports waste heat from the powerconversion devices to radiator panels

o A solid-state power conversion package (SiGe/GaP) thaitransforms the reactor thermal power into electricity

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Schedule

o The largest and most visible component of the system, a wasteheat radiator that dissipates the heat to space (beryllium orcarbon-carbon matrix structure with titanium/potassium heatpipes).

In the fully deployed configuration, the 100 kWe power generation unitmeasures about 12 meters from the tip of the reactor to the tip of theradiator.

The SP-100 reactor development program started tenfold years ago, but theprogram has recentely been terminated and 20 types of advanced tech-nologies are now under transfer to the industry.

SP-100 Unit Data

Thermal power

Gross power generated

Peak reactor outlet temperature EOL

HT loop AT

HT loop mass flow

Peak radiator inlet temperature

HR loop AT

HR loop mass flow

Average radiator surface temperature

Radiator block body area

Radiator physical area

PC thermopile area

T/E leg length

2.3 MWt

105.3 kWe

1375 K

92 K

5.9 kg/s

840 K

48 K

10.4 kg/s

790 K

94 m2

104 m2

5.5 m2

0.55 cm

References BAILEY, H S et alSP-100 space nuclear Power Technology Requirements for kW- to MWsystems.Transaction American Nuclear Society, 1992, Space nuclear power, p 21-22.AIP Conference Proceedings No. 324,1995, January 8-12, USA, NM,Albuquerque.

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Figure 15SP-100: Layout.

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16:1(3)

16

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

ERATO

SNR, power source

LMR/HTGR

Very small power/20 kWe

CNES/CEA, France

Studies

Since 1982, France has be200 kWe space nuclear power systems adapted to the Ariane V launcher fortwo reference missions: space-based radar and electric propulsion.

An ERATO space nuclear power system concept consists of:

o A nuclear reactor cooled by gas or liquid metal which is theheat source of the reactor

o An energy conversion system to produce electric power fromnuclear heat carried by the coolant

o A radiator which is the heat sink of the system and radiates thenuclear heat not transformed into electricity in space

o A shield to protect payload and sensitive equipment fromneutron and gamma photon irradiation

For the 20 kWe ERATO plant the following concepts were considered:

o A gas-cooled epithermal reactor with a UC2 fuel particle bedand superalloys (HRA) as structural materials (Hastelloy X forinstance) with a core exit gas temperature of 820 °C

0 A sodium or sodium potassium (NaK) cooled UO2 fuel fastneutron reactor with 316 stainless steel material operating at amaximum temperature of 700 °C

The 20 kWe ERATO reactor plant considered generator employs a Braytoncycle energy conversion system with He-Xe working fluid. The gas is heated

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16:2(3)

Schedule

through a liquid metal-gas heat-exchanger in liquid metal-cooled reactorsystems and directly in the reactor core in the gas-cooled reactor system.

One or two Brayton turboelectric converters (gas turbine, compressor,alternator) are intended in the 20 kWe unit to permit a back-up operation atquasi nominal power in the event of partial unavailability conversion unit andto improve the global system reliability.

Considering CNES and CEA experience in the space and nuclear fields, itwould be possible to develop in France a space nuclear power system if itwas decided. But up to now no mission requiring such electric power hasbeen envisaged in France, or in Europe. Nevertheless the development ofhigh electric power system which represents a new milestone in the conquestand utilization of space constitutes a very long process.

ERATO Unit Data

ReactorThermal powerUranium massFuelStructural materialPrimary coolantPressureCore inlet/outlet temperatureControl drums (Be + B4C)Safety rods

HOkWt70 kgUO2S steelNaK2.5 bar577/682 °C7(B4C)

125 kWt137 kgUC2

SuperalloyF Xe80.0 bar511/820 °C19(B4C)

Shield

Intermediate heat exchanger

B4C+LiH B4C+LiH

Gas inlet/outlet temperature

Brayton turboelectric converterTurbine inlet temperatureTurbine inlet pressureCompressor inlet temperatureCompressor pressure ratio

RadiatorPowerInlet/outlet temperatureArea

430/670 °C

670 °C9.0 bar59 °C2.06

80kWt236/53 °C86 m2

430/670 °C

819°C78.0 bar119°C2.16

104kWt368/119»C54 m2

Global system efficiency 0.19 0.16

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16:3(3)

References CARRÉ, F, DELAPLACE, J, PROUST; E and TILLIETT, ZSpace nuclear power system studies in France and envisaged missions.Proceedings Nuclear Power Engineering in space, Obninsk, RF, 15-19 May,1990.

SAFETY RODS REFLECTOR

REACTOR CORE

Figure 16ERATO: Reactor block (Ud/HeXe/HRA 850 °C).

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17

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

LMCPBR

SNR, power source, moon-base

LMR/HTR

Small power/10 MWt

Toshiba/JAERI, Japan

Studies

A concept of a power 3eneratini

Schedule

bed reactor with a thermal output of 10 MWt has been studied by theToshiba & JAERI team for use in a moon base. This reactor concept is basedon the HTGR fuel technology and the liquid metal cooling technology. Thedesigned system consists of 71 lithium cooled pebble beds, a fast spectrum,drum-controlled reactor, a liquid metal steam generator, a turbogenerator,and a panel heat rejection system.

The secondary coolant is potassium which is vaporized in a steam generatorand rotates a turbine-generator. Both coolants are circulated by electro-magnetic pumps.

The LMCPBR plant used the potassium Rankine cycle. A reactor vesselcontains an integrated potassium boiler for reducing the total mass as muchas possible. As a reactor structure material, Nb-Zr alloy is used consideringthe material strength at high temperatures. The total system mass is about17 tonnes.

No LMCPBR plant has been ordered yet. A conceptual design study is underconsideration.

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17:2(3)

LMCPBR Unit D a *

Therma! power

Reactor coreDiameterHeightUranium massC/U ratio

Structural material

ReflectorAxial- Material- ThicknessLateral- Material- Thickness

Control drumDiameterNumber of drumsPositionMaterial

Primary coolantMaterialInlet temperatureOutlet temperatureFlow rate

lOMWt

1.1m1.1m503 kg35.8

Nb-Zr alloy

BeO pebble (3 cm diam)25 cm

Be28 cm

24 cm18In lateral reflectorB4C (80 % 10B) and Be

Lithium-71 123 K1203 K3.9 nvtymin

References HIROMICHI, N et alA liquid metal cooled pebble-bed reactor for a moon base.Proceedings ICENES '93,1993,20-24 Sept, Makuhari, Chiba, Japan,p 353-363.

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Fuel charger

203K

Lithium

Reactor

1123K

1995-02-13

1159KBypass ivalve X

Steamgenerator

1123K

Potassium

Panel radiatorcondenser

823KSurfacecooler

EM pump

Figure 17LMCPBR: Line diagram.

833K

EM pump

Cold trap

Flow control valve

Cold trap

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18

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

Rover/NERVA

SNR, propulsion

Prismatic solid core

Medium power/367

Westinghouse/LAN

Basic design

The basic Rover/NI

Schedule

1.32 m long by 0.89 m in diameter supported in compression byregeneratively-cooled Inconel tie-tubes passing through unloaded (nouranium) support elements. The fuel elements are 19 hexagons 1.91 cmacross flats. Fuel/ support element flow passages are nominally 0.254 cm indiameter. A radial beryllium reflector, 11.43 cm thick, houses 12 rotatablecontrol drums that provide neutronic control. Fuel loading is varied radiallyto partially flatten power, and flow orifices at the inlet of each coolantpassage match the flow to the local power. An aluminium inlet-end supportplate, a shadow shield, and an aluminium pressure vessel are other majorreactor components. Core diameter ranged between 53.3 cm (PEWEE 1)and 140 cm (PHOEBUS 2A). The flow passage diameter and fuel elementsize may, in principle, be changed. However, the element dimensions aregoverned by a complex set of performance and fabrication factors and thepresent dimensions arc considered to be near-optimum.

More than 50 000 fuel elements have been fabricated and tested during the17 years of the Rover/NERVA program. The main efforts have been devotedto fuel fabrication particles and to insertion in graphite matrix.

In 1973 the Rover/NERVA program was terminated. The futuredevelopment of nuclear propulsion systems can be based on the technologyand experience of the past efforts.

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18:2(3)

Rover/NERVA Unit Data

Type

Thrust

Specific impulse

Thermal power

Turbo pumppower

Turbo pumpspeed

Pump dischargepressure

Engine flow rate

Chambertemperature

Chamberpressure

References BOHL, R J et al

XE

245 kN

6970m/s

1 141 MWt

5.1 MWt

22 270rpm

6.80 MPa

35.9 kg/s

2 270 K

3.86 MPa

Nuclear rocketry review.Fusion Technology, v 20, Dec, 1991,

NERVA

337 kN

8 085 m/s

1566 MWt

23 920 rpm

9.36 MPa

41.9 kg/s

2 360 K

3.10 MPa

p 698-670.

Smallengine

72 kN

8 575 m/s

367 MWt

0.9 MWt

46 950 rpm

6.03 MPa

8.5 kg/s

2 695 K

3.10 MPa

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1995-02-13

REFLECTOROUTLETPLENUM

REFLECTOR

FLOY.

ID DO*-«EEND PLENUM

Fig. 5

Figure 18Rover/NERVA: Reactor layout.

PRESSURE

VESSEL

CORE INLET

PLENUM

LATERALSUPPORT SYSTEM

REFLECTORINLET PLENUM

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19

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

NPPS

SNR, power source/propulsion

Epithermal solid core reactor

Medium power/1 200 MWt

RRC-KI.RF

Studies

A Nuclear Power Propulsion S

Schedule

output of 1 200 MWt has been developed by RRC-KI for a manned Martianmission.

The advisability of using a nuclear power source in a space engine system isdetermined by the possibilities of die nuclear rocket engine realizing diespecific impulse of 9 000 m/s in the propulsion mode using hydrogen as dieworking fluid, and also widi a good diermal power capability and masscharacteristics fcr die nuclear power engine system widi a closed cycle in diepower mode. One of die possible variations of die nuclear power enginesystem being considered is based on die concept of die nuclear rocket engineand die turbine electric engine converters.

In die engine mode, hydrogen is used as a coolant and is heated in diereactor to a temperature of 2 600 - 2 900 K. In die power engine mode(Brayton cycle) die working fluid is a mixture of helium and xenon. Researchresults have shown that die most acceptable system is an intermediate(epithermal), solid-core reactor of uranium carbide, niobium and zirconium,widi a reflector made of beryllium.

This concept was based on basic theoretical research conducted in die formrrSoviet Union and abroad, and also on neutron research on critical assemblies,power tests on reactor facilities, and on prototype reactor installations. Nocurrent development

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19:2(3)

NPPS Unit Data

Power propulsion scheme

Module reactor configuration

Nuclear fuel

Propulsion modePropellantThermal powerSpecific impulsesCoolant flow rateCoolant outlet temperature

Power mode

Coolant

Energy conversion system

Electrical powerCoolant maximum temperature

Radiation environmentCrew exposure to NPPS radiationEstimated doses in the instru-mentation compartment:- y-radiation- Fast (<0.1 MeV) neutrons influenceRadiation energy release in hydrogentanks during propulsion mode

NPPS total mass

3 - 4 modules

Channel housing

Solid solution ofuranium, niobium,zirconium carbides

Hydrogennot over 1 200 MWt(8 - 9) • 103 m/s22 - 25 kg/s2 600 - 2 900 K

Xenon and heliummixture (I - 3 % He)Turbomachine(Brayton cycle)50-200kWe1200 K

Not over 10 rad

Not over K^radNot over 1013n/cm2

Not over

50 - 70 Mg

References IVANOV, A ANuclear Power Propulsion System Concept for a Manned Martian Mission.Space Nuclear Power system, 1989, Edited by El-Genk, M S andHoover, M D, Orbit Book Company, Malabar, FL, 1992, v 11 p 475-477.

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19:3(3)

RadiatorWorking fluid tanks

ITPPS

Figure 19NPPS: Schematic.

Possible

crew

location

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1995-02-13

JO

Subcritical Accelerator-Driven WasteTransmutation and Energy GenerationSystems

Introduction

Acceleiator-driven waste transmutation and energy generation have gainedrenewed interest through studies on different long-term waste managementoptions. The possibility of a future global power production system that isfree from the discharge of long-lived radioactive products (TRU and fissionproducts) is under consideration in France, the USA, Japan, Russia, Swedenand Switzerland.

The initial steps deal with the analysis of neutron economy to evaluate andcompare the transmutation potential of different nuclear power systems aswell as losses in the reprocessing, partitioning and fuel fabrication. Thepotential decrease of radiotoxicity is also an important subject for thesestudies.

It has been suggested that accelerators could be a part of advanced systemsin order to produce energy and additional neutrons to transmutate wastenuclides or breed fertile material to fissile nuclides (Th-232 to U-233 orU-238 to Pu-239).

The blanket elements associated with an accelerator are similar to a fissionreactor core concerning the neutron fluxes, heat generation and the radio-active content. Such a system has the advantage of operating in a subcriticalmode. Nevertheless problems with loss of coolant are the same because thedecay heat is similar and related to the power production.

From the economical point of view it has the disadvantage of the additionalinvestment of the accelerator and that a significant fraction (10 - 30 %) ofthe produced electricity will be required to operate it.

Such a hybrid system with a high transmutation rate will be able to supportthe burning of TRU and long-lived fission products in the final stage ofnuclear fission power assuming that the toxicity of spallation products isnegligible.

Substantial technical development will be necessary to achieve an acceptablesolution for the containment, the interface between the accelerator vacuumsystem and the target/blanket as well as accelerator parameters.

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J:2(2)

References

Proceedings International Conference on Accelerator-DrivenTransmutation Technologies and Applications, Las Vegas, NV,US, July 25-29,1994.

Future Nuclear Systems: Emerging Fuel Cycles & WasteDisposal Options.Proceedings Global '93 International Conference, Sept 12-17,1993, Seattle, Washington, US.

Accelerator-Driven Transmutation Technology, 1994, October,N2.

SEKIMOTO.HetalNuclear system research and development in the next centurytoward an equilibrium state.Global "93, Sept 12-17, Seattle, Washington, Proceedings,Vol l .pp 283-287.

SALVATORES, MetalA global physics approach to transmutation of radioactivenuclei.Nuclear Science and Engineering, 1994, V 116 (1), pp 1-8.

IGNATIEV, V and SUBBOTINE, SOn the possible contribution of MSRs to nuclear power atfuture stages of its development.Proceedings 4th NSI Conference, 28 June - 2 July, NizniNovgorod, RF, 1993.

SKÅLBERG, M and LILJENZIN, J-OPartitioning and transmutation, the state of the art.Nuclear Engineering International, 1993, February, pp 30-33.

DEVELL, L, IGNATIEV, V and SUBBOTINE, SSafety aspects of waste management and disposal for presentand future nuclear options.Proceedings 5th NSI Conference, 27 June - 2 July, Obninsk,RF, 1994.

Overview of Physics Aspects of Different TransmutationConcepts.1994, OECD, Report NEA/NSC/DOC(94)11.

Development Technology to Reduce Radio-active Waste MayTake Decades and be Costly.1994, USA, Report GAO/RCED-94-16.

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1995-02-13

J l

Format

Tide


Reactor type

Power output

Organization (name)

Development status

Description

Description

Energy Amplifier

NPP

SADS/PWR (HTGR)

Small power/200 MWt

CERN, Switzerland

Studies

The concept of energy

Schedule

of an accelerator-induced nuclear cascade with the 200 MWt thermal outputis presented by CERN. The beam of high energy particles is directed into afuel material target Thorium is used as breeding fuel. The neutrons therebyproduced are slowed down by the moderator medium and multiplied undersubcritical conditions by the breeding and fission processes.

The Energy Amplifier design consists of two main separate parts: mediumcurrent (1 -10 mA), medium energy (1 GeV), proton accelerator feeding asubcritical assembly consisting of Th and moderated media (H2O orgraphite).

Under conditions of moderate neutron flux (1014 n-cnr^s-l), the con-centration of fissile U-233 which is bred from Th is stable at about 1.3 %.The U-233 produces energy fission and is continuously regenerated in situwithout resorting to any chemical separation. It is shown that the energyproduced is several times larger than that required to power the protonaccelerator.

This system will result in very small quantities of Plutonium and higheractinide waste. Based on the abundant and inexpensive resource which isnatural Thorium, this system can be built using present day reactortechnology according to the proponents.

No Energy Amplifier reactor design has been ordered yet A conceptualstudy is in progress.

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1995-02-13

Jl:2(3)

Energy Amplifier Unit Data

References

Accelerator type

Gain

Beam kinetic energy

Average current

Beam input

Energy amplifier type

Fuel cycle

Fuel type

Neutron flux

Capacity

Coolant temperature,inlet/outlet

Coolant pressure

RUBBIA. C et al

LINAC

40

0.8 GeV

6.25 mA

5.0 MW

PWR

Th-233u

TnO.-pins

1014n-cm-2.s-l

200MWt/60MWe

291 °C/322 °C

154 bar

Cyclotron

40

0.8 GeV

6.25 mA

5.0 MW

HTGR

Th-233u

ThOyThC pebble withPb/Pb-Bi target

lO^n-cm-Zs-1

200 MWt/80 MWe

435 °C/710 °C

58 bar

An energy amplifier for cleaner and inexhaustible nuclear energy productiondriven by a particle beam accelerator.1993, Geneva, Switzerland, Rep. CERN/AT93-AT93-47(ET).

RUBBIA, CThe Energy Amplifier Concept: A solid-phase, accelerator driven subcriticalTh/233ij breeder for nuclear energy production with minimal actinide waste.Proceedings International Conference on Accelerator Driven TransmutationTechnologies and Applications, Las Vegas, NV, July 25-29,1994.

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STUDSVIK ECO & SAFETY AB STUDSVIK/ES-95/10 Jl:3(3)

1995-02-13

Coritrol bars

Supportstructure

Core supportbarrel Fuel handling

and control bars

£ Magnetic { S S t ^ i S l S is quadrupdepi;ga^^j|||jpi-" tenses PS€^^€*^S

->' :;ä Beam monnonng

-. : Beam monitoring L i t ä^Cs

. Beam monaonngdevice

Figure JlEnergy Amplifier: Schematic (PWR-blanket version).

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1995-02-13

J2:i(3)

J2

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

PHOENIX

TWS

SADS/LMR

Large power/3 600 MWt

BNL, US

Studies

Three accelerator-driven

Schedule

in BNL. BNL has focused the concepts just on the minor actinides and long-lived fission product incineration.

The PHOENIX concept assumes a large Linac that can produce a 104-mAbeam of 1.6 GeV protons. A multiple module concept was developed for thePHOENIX subcritical lattice. Each module resembles the core of the FastFlux Test Facility (FFTF) with the minor actinides formed into oxide fuelrods, replacing the uranium and plutonium in the FFTF fuel. The fuel rodsare cooled using liquid sodium and are bundled into 217 pin assemblies, with124 such assemblies making up a 450 MWt target module. One to eight ofthese target modules are aligned in front of the proton beam, depending inpan on how much of the "fuel" is available at any given time. A current of104 mA is sufficient to drive with k = 0.9 subcritical lattice at 3 600 MWt.The proposed machine, based on the described PHOENIX concept, wouldtransmutate the minor actinides and the iodine produced by 75 LWRs andwould generate usable electricity of 850 MWe.

The second BNL concpet with multi-segmented cyclotron injects 2.0 - 5 mAbeam of 1.5 GeV protons into the lead target located at the center of theMOX-fuelled fast reactor at a slightly subcritical mode.

The third BNL concept uses a multi-segmented cyclotron to inject 4 - 8 mAbeam of 1.5 GeV protons into the lead spallation target located at the centerof the fast spectrum core, which is loaded with nitride coated particle fuel.

The conceptual design study is not completed and the R&D program isunder way.

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MOX Fuel Core System Unit Data

Accelerator

- Type

Particle

Energy

Current

Target

Equivalent diamater

Equivalent height

Target material

Cooling material

Subcritical core

Equivalent diameter

Equivalent height

Material composition

Fuel/Clad+Structure/Coolant/Moderator

Chemical form of fuel

Materials of coolant and moderator

Averaged fresh fuel compositionU/Pu/Np, Am,Cm/LLFP

Isotopic composition of Pu238PU/239PU/240PU/241 P u / ^ p u

Multi-segmentedcyclotron

Proton

1 500 MeV

2.0 - 5.0 mA

10 cm

75 cm

Pb

He

180 cm

93 cm

35/34/41/0 v/o

Oxide

Na / -

73/22/5/0 w/o

0/58/24/14/14 w/o

Isotopic composition of MA237Np/24lAnV^2Am/243Am/243Cm/244Cm 53.6/23.1/0/17.4/0/5.9 w/o

Averaged composition of LLFP99TC/129!

System characteristics

Effective multiplication factor:

Thermal power in core

Power density: average

94/6 w/o

0.98 - 0.99

700 MWt (280 MWe)

930W/cm3

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J2:3(3)

References VANTUYLE,GJetalAccelerator-driven subcritical target concept for transmutation of nuclearwastes.Nuclear technology, 1993, Vol 101, January, pp 1-17.

VANTUYLE,GJAccelerator-driven target technologies under development at BNL.Proceedings International Conference on Accelerator-Driven TransmutationTechnologies and Applications, Las Vegas, NV, US, July 25-29,1994.

850 MW(electnc)

410 MW(electric) toAccelerator

LinearAccelerator

Input per Year2.6 tonnes of

Minor Actinides580 kg Iodine

BeamExpander

104 mA1.6-GeVProtons

Output per Year1.05 tonnes Fission Product

1.55 tonnes Plutonium580 kg Xenon

Figure J2PHOENIX: Schematic.

SteamGeneratorSystem

1260MW(electric)

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J3:i(3)

J3

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

JAERI-TPC

TWS

SADS/MSR (LMR)


JAERI, Japan

Studies

In Japan the high-level rat

Schedule

OMEGA project (Options for Making Extra Gains from Actinides andFission Products) proposed by The Science and Technology Agency (STA)in 1988. Three types of accelerator-driven transmutation systems are underdevelopment at JAERI.

In the JAERI-TPC concept accelerator injects proton beam of 1.5 GeV,40 mA into the tungsten target located at the center of the reactor corewhich is leaded with minor actinides and other long-lived nuclides. Thissystem is capable of increasing about 260 kg per year of minor actinides.

The core design of the JAERI-TPC concept is based on solid-state pin-bundle type fuel elements with sodium cooling. It offers some advantageslike high power density, and high temperature operation under normalpressure.

A study is also in progress on the system with 1.5 GeV, 25 mA proton beamaccelerator and hard spectrum chloride molten salt assembly which isintended for transmuting not only minor actinides, but also fission products.This concept would be a continuous processing system.

In the third concept accelerator injects 1.5 GeV proton beam of 20 mA intoeutectic alloy (Np-Pa-Ce-Co-Tc) target core with graphite blanket, which iscooled by molten salt fluorides.

The conceptual design phase is not completed.

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1995-02-13

JAERI-TPC Unit Data

Proton beam CUITCM

Actinide inventory

keff

Number of neutrons per incident proton

Number, of fissions(>15 MeV)« 1 5 MeV)

Neutron flux

Mean neutron energy

Bumup rate

Thermal outputFuelTungstenTotal

Power densityMaximumAverage

Linear power ratioMaximumAverage

Coolant temperatureOutlet maximum

average

Maximum temperatureFuel center

surfaceGad inside

outside

39 mA

3 610 kg

0.89

40n/p

0.45 f/p100 f/p

4 x 1015 n/cm2/s

690 keV

250 kg/year

800 MWt20MWt

820 MWt

930 MW/m^400 MW/m3

61 kW/m27 kW/m

437 °C430 °C

890 °C548 °C528 °C484 »C

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J3:3(3)

References Mfg. of Accelerator Begins for PNCs Transmutation Research.Atoms In Japan, Janurry 1994.

KAWATA, T, YOSHIDA, H and HATTA, H"Overview of R&D Program on Nuclide Partitioning and Transmutation(OMEGA) in Japan".NRC Symposium on Separation Technology and Transmutation Systems(STATS), Washington, USA, January 13-14,1992

Primary Na In '= j

5m4

32

I

L_ Q

Figure J3JAERI-TCP: Schematic.

Primary Na Out

Proton Beam

Beam Window

Tungsten Target

Reactor Vessel

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STUDSVnC ECO & SAFETY AB STUDSVIK/ES-95/10 J4:i(3)

J4

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

ATW/ABC

TWS/MSR (LMR)

SADS


LANL.US

Studies

In recent years, acceleratplutonium utilization (ABC), energy production (ADEP) and transmutationof waste (ATW) systems stimulated interest in LANL as a technologicallypossible method.

The most promising design option for an advanced ATW/ABC/ADEPconcept uses 0.8 GeV, 90 mA proton beam LINAC; a fluid lead/or lithiumtarget and a multiplying blanket which is graphite-moderated and moltensalt-cooled. The systems operate at a value for k ^ of between 0.9 and 0.95.The ADEP system burns commercial waste at this rate and also generates2 250 MWt in the Th-233U cycle while also burning the waste origniatedfrom thorium. On-line separation would be used for fuel processing. Thelithium target offers efficient neutron production (estimated at 28 neutronsper 800 MeV deuteron) and low absorption of blanket neutrons because ofits small size and presence of the fission product transmutation regionsurrounding it This target produces a very small amount of heavy spallationproducts.

The high intense thermal flux 1016 n/(cm2s) in the blanket of the protonaccelerator allows for the effective transmutation of main waste actinideNp-237 through conversion into fissile Np-238, because in a thermal flux ofa typical power reactor Np-237 is converted into neutron poison in lowintense neutron irradiation (~1014 n/(cm2 s))

The principle features of the ATW/ABC/ADEP concept are the following:

o Subcritical operation

o Low inventory of actinides and fission products: high thermalneutron flux with no dilution in external heat exchangers. Non-

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J4:2(3)

radioactive coolant No spallation products in target. Lowpressure

o

c

Negative temperature and void reactivity coefficient

Passively safe in the event of a loss of coolant accident. Largethermal inertia and good heat conductivity. Molten salt can bedrained away without external intervention

Proliferation-resistant fuel cycle. Uranium enrichment lowerthan 20 %. Thorium breeding without reprocessing

ci dule The conceptual study is not completed.

References BOWMAN, C DOverview of tue Los Alamos accelerator-driven transmutation technologyprogram.Proceedings International Conference on Accelerator-Driven TransmutationTechnologies and Applications, Las Vegas, NV, July 25-29,1994.

KONDE, HThe Los-Alamos concept for accelerator-driven energy production andtransmutation of nuclear waste.Rep Uppsala University/Studsvik Eco & Safety AB, UU-NF 94/112, 1994,I/ecember.

90 mi 800 MeVAccelerator

One

Spent Fuel

Assembly

It and URemoval

2i

Storage

750 MWt Actinic* Waste2250 MWt Th ' » u Cycle

Stor agej

ATW target/blanket

3000 MWt

AcunideSeparator

Actmides

IFissionProd 7 5%

j TransmutedFissionProduce

Fission ProrJuc;;Storage

Stable & Shortlived

FissionProducts

FissionProductsSeparator

Figure J4 (2 pages)ATW/ABC: Schematic flow sheets.

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STUDSVIK ECO & SAFETY ABSTUDSVIK/ES-95/H)

1995-02-13

J4:3(3)

Fual-Salt Motor

Target EM Pump»>

Upper Axial Ref lectoi

Fuel-Salt Pui

Sparge BubbleInjector/Separator

Radial Reflector'

—•Downcommer Baffle!

Lower Axial Reflects

ProtonBeam , Target Secondary Coolant

Target Heat Exchanger

Secondary Coolant

Primary Heat Exchanger

Secondary, Containment

Vessel

• Target Window

PrimaryContainment

Vessel

Secondary Coolant

Downcommer Baffle

Radial Reflector

Molten Pb Ta

Fuel Salt Pump

Downcommer Baffle

Primary Heat Exchanger

SecondaryContainment

VesselPrimary

ContainmentVessel

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J5:i(3)

J5

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

AMSB (Accelerator Molten Salt Breeder)

TWS.NPP

SADS/MSR

Medium power/1 000 - 2 100 MWt

Tokay University, Japan

Studies

The AMSB plant design composed of a pi

Schedule

ww victor/J5 ea

fluoride target system as a fissile breeder is considered at Tokay University(Japan).

AMSB is composed of three parts: an 1 GeV-300 mA Linac, a molten-fluoride target/blanket system, and a heat transfer and electric powerrecovery system. The size of the target LiF-BeF2-ThF4 salt bath is 4.5 ~ 5 min diameter and 7 m in depth. To keep them smaller in size, a comparativelyslow proton of 1 GeV was chosen. The inside of the breeder vessel made ofHastelloy N (Ni-Mo-Cr alloy) is covered by thick graphite blocks immersedin salt. The target salt is introduced from die upper part forming a vortex ofabout 1 m in depth of salt. The proton beam is directly injected into the off-centered position near the vortex bottom, saving the neutron leakage andimproving the heat dissipation.

The key features of the AMSB concept are the following:

o The target/blanket system is subcritical, and is not affected byradiation damage

o The heat removal could be managed by the dynamic salt, which

is diluted in heavy isotopes

o The shuffling is automatic

Unsolved technological problems seem to be:

o High current Linac development

o Proton beam injection part engineering

The conceptual study has not been completed.

STUDSVIK ECO & SAFETY AB•TYAB

AMSB Unit Data

Type

Proton beam

Target salt

Salt temperature,inlet/outlet

Molten salt loading

Neutron production perproton

U-233 production

Fissile inventory(doubling time)

Incineration ratefissionspallation

Thermal output

Electric output

Linac consumption

STUDSVIK/ES-95/10 J5:2(

1995-02-13

Standard type

IGeV, 300 mA

7LiF-BeF2-ThF4

580°C/680°C

2431

25-40

(0.57-0.92) t/year

(0 - 0.5) t(0-1 yr)

-46 kg/yr-46 kg/yr

1000 - 1500 MWt

430 - 650 MWe

600 - 700 MWe

High-gain type

IGeV, 300 mA

7LiF-BeF2-ThF4-233UF4

580 °C/680 °C

1261

36-58

(0.82-1.33) t/year

3.41( 2 - 4 years)

-174 kg/yr-46 kg/yr

1400-2100 MWt

600 - 900 MWe

600-700 MWe

References FURUKAWA, K et alFlexible thorium .nolten-salt nuclear energy synergetics.Proceedings Potential of small nuclear reactors for future clean and safeenergy sources, Tokyo, Japan, 23-25 October, 1991, pp 13-22.

FURUKAWA, K and CHIGRINOV, SPlutonium (TRU) transmutation and 233U production by single-fluid typeaccelerator molten-salt breeder (AMSB).Proceedings International Conference on Accelerator-Driven TransmutationTechnologies and Applications, Las Vegas, NV, July 25-29,1994.

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proton beam neutrondifferentialpumping

stopper

graphite

graphite vortei ^ \

vapor trop K ^ e T s o T * .Ktnin flowv ^V(/ f f t

WWW \ \ rrr-7" /

graphite

i : : —vortex

1^-w—i^lTE 53Vprocessfacilitiesover flowline " ^

pump

mainpump heat

exchangerprimary loop

tank

Hostelloy N

graphite

target salt

Figure J5AMSB: Schematic.

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1995-02-13

J6:i(2)

J6

Format

Title


Reactor type

Power output

Organization (name)

Development status

Description

Description

BBR (Breeder-Burner Reactor Subcritical System)

TWS

SADS/MSR


CEA, France

Studies

Hybrid system Breeder-Burner Reactor Subcriticalinvestigated at CEA, mainly from a conceptual point of view, in order toassess its potential to transmute radioactive wastes (mainly long-lived fissionproducts, LLFP) and its potential to insure a minimal long-term radiologicalrisk related both to the fuel inventory inside the system and to the full fuelcycle (mass flows, reprocessing, transport waste disposal).

The system under study uses:

oooooo

The Th cycleHigh fuel burn-up (-45 % of heavy atoms)Fast neutron spectrum and a flux level —2-1015 n/s/cm2

Subcriticality in the range of keff =0.85Neutron balance improvements (e.g. leakage decrease)Fuel breeding (to compensate for reactivity swing)

The system has an external neutron source via proton beam (1.5 GeV,270 mA), and the fuel of the subcritical reactor acts as target and is made ofmolten salts (Th + 3 % Pu-239) Q2 + Pb C^ (where Pu-239 can be replacedby 5 % TRU), with the addition of Tc-99. The choice of molten saltchlorides is not essential, but taken as an example of a high bum-up con-figuration. The electric power of the system is 2 GWe, with a maximumspecific power of -70 W/cm3. The fuel lifetime is 50 years with reloadingintervals of 10 years.

The following rates of transmutation are obtained: Pu-239: 0.55 t/GWe/yr;Tc-99: 0.062 t/GWe/yr

The decrease of the waste toxicity as compared to the feed fuel is a factor of-40 for "long" time intervals (102 - 10 6 years) and - 1 500 for "short" tiemintervals (102 -104 years).

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J6:2(2)

Schedule Conceptual studies are under way now.

BBR Unit Data

Accelerator

- Type

- Energy

- Current

Target/Core

- Diamatcr

- Height

- Cooling material

- Averaged fresh fuel compositionTh/Pu/^Tc weight ratios

- Isotopic composition of Pu

- Isotopic composition of LLFP

System characteristics

- Ktff

- Core thermal power

- Power density: average/max

- Neutron flux averaged in core

- Fuel (target) dwelling time

- Bumup reactivity swing

- Bumup

- Reloading interval

Proton LJNAC

1.5 GeV

270 mA

495 cm (all core)

600 cm

60 (Th+3% 239Pu)Cl2 -40% PbCl3 mol %

35/1.1/1.85

100 % 239Pu, no TRU

100%"Tc

0.85

5 000MW1

40/70 W/cm3

2 x 1015 n/cm2/s

50 years

0.003 %6k/k

45 % heavy atom

10 years

References SLESAREV, I, SALVATORES, I and UEMATSU, MWaste transmutation with minimal fuel cycle long-term risk.Proceedings International Conference on Accelerator-Driven TransmutationTechnologies and Applications.Las Vegas, NV, July 25-29,1994.

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Appendix 1:1

List of Abbreviations and Glossary of Terms

A

ABB

ABWR

Accelerator

Actinides

Active component

Active system

ADP

ADS

AEC

AECL

AGR

ALARA

ALMR

ALSEP

ALWR

AMSB

amu

ANL

AP

APWR

ATR

ATWS

Availability Factor

AWTS

Asea Brown Bovcri

Advanced Boiling Water Reactor, see BWR

A device that increases the velocity and energy of charged particles suchas electrons and protons; also referred to as a particle accelerator. In a"linear" accelerator, particles are accelerated in a straight path

The elements with atomic numbers above 88 (actinium, element 89). Theactinides series includes uranium, atomic number 92, and all the man-made transuranic elements

Any component that is not passive is active

Any system that is not passive is active

Advanced Double Pool reactor

Automatic Depressurization System

(ILS) Atomic Energy Commission

Atomic Energy of Canada Limited. A designer/supplier of nuclearreactors

Advanced Gas-Cooled Reactor

As low as reasonable achievable

Advanced Liquid Metal Reactor

Apollo Lunar Scientific Experiment Package

Advanced Light Water Reactor, see LWR

Accelerator Molten Salt Breeder

Atomic mass unit

Argonne National Laboratory, US

Advanced Passive

Advanced Pressurized Water Reactor, see PWR

Advanced Thermal Reactor: A heavy water moderated, light water-cooled reactor

Anticipated Transients Without Scram

The availability factor of a nuclear unit or station is the ratio of timewhen energy can be produced to the total time

Accelerator-driven Waste Transmutation System

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1995-02-13

Appendix 1:2

B

Base Load

BDA

BMDO

BN

BNL

BOL

BOM

BOP

Breeder Reactor

Breeding Ratio

BWR

The minimum load produced over a given period. A station used forbase load is a station that is normally operated when available to providepower to meet the minimum load demands

Beyond Design Accident

Ballistic Missile Defence Organization, US

Russian version of sodium cooled Fast Breeder Reactor

Brookhaven National Laboratory

Beginning Of life

Beginning Of mission

Balance Of Plant

A nuclear reactor that produces more fissile material than it consumes.In fast breeder reactors, high-energy (fast) neutrons produce most of thefissions, while in thermal breeder reactors, fissions are principally causedby low-energy (thermal) neutrons. See Fast reactor

The conversion ratio when it is greater than unity. A high breeding ratioresults in a short doubling time

Boiling-Water Reactor. A light-water reactor that employs a directcycle; the water coolant that passes through the reactor is converted tohigh pressure steam that flows directly through the turbines

CANDU

Capacity Factor

CAREM

CCV

CDA

CE

CEA

CEC

Charge of a reactor

CHP

CANadian Deuterium Uranium reactor: A type of heavy water reactor

See Load Factor

Conjunto Argentina de Reactores Modulares

Cooldown Control Valve

Control Drum Actuator

Combustion Engineering

Commissariat å 1'Energie Atomique

Commission of the European Communities

The fuel placed in a reactor

Combined Heat and Power

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1995-02-13

Appendix 1:3

Conversion as used inreactor technology

Conversion, chemical

Converter reactor

Coolant

COPUOS

Critical

CRU

CWCS

The ratio between the number of fissile nuclei produced by conversion tothe number of fissile nuclei destroyed. If the ratio for a given reactor isgreater than one, it is a breeder reactor, if it is less than one, it is aconverter reactor

The operation of altering the chemical form of a nuclear material to aform suitable for its end use

A reactor that produces some fissile material, but less than it consumes

The medium in a nuclear reactor which removes heat from the reactorcore where fission occurs and heat is produced, and transfers it tosystems which convert the heat into steam

(UN) Committee On Peaceful Uses of Outer Space

Capable of sustaining a nuclear chain reaction

Combined Rotating Unit

Containment Water Cooling System

D

DARPA

DBA

DEC

Decay Heat

DIPS

DNHPP

DOE

E

Defense Advanced Research Project Agency

Design Basis Accident

Direct Energy Conversion

The heat produced by the decay of radioactive nuclides

Dynamic (Radio)isotope Power System

Decentralized Nuclear Heating Power Plant

US Department of Energy

E-MAD

EASEP

ECCS

EFR

EHRS

EM

EMP

Engine Maintenance, Assembly and Disassembly

Early Apollo Scientific Experiment Package

Emergency Core Cooling System

European Fast Reactor

Emergency Heat Removal System

Electromagnetic

Electromagnetic Primary Pump

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Appendix 1:4

EMP

Enriched fuel

Enrichment

Fast neutrons

Fast reactor

Electromagnetic Pump

See Fuel, enriched

i) the fraction of atoms of a specified isotope in a mixture ofisotopes of the same element when this fraction exceeds that inthe naturally occurring mixture;

ii) any process by which the content of a specified isotope in anelement is increased

EOL

EOM

EOT

EPA

EPR

EST-1

eV

EVA

EWST

F

Fail-safe

End Of life

End Of Mission

Earth Orbital Terminal

(U.S.) Environmental Protection Agency

European Pressurized Reactor

Engine Test Stand Number One

Electron volt

Extravehicular Activity

Emergency Water Storage Tank

The term describes the behaviour of a co

Fault-/error-tolerant

FBR

FDA

FEPS

failure (either internal or external). If a given failure leads directly to asafe condition, the component or system is fail-safe with respect to thatfailure

See Neutron, fast

A nuclear reactor in which no moderator is present in the reactor core orreflector. So the majority of fissions are produced by fast neutrons. If afertile species is present in the fast reactor core or in the blanketsurrounding the core, it will be converted into fissile material by neutroncapture. When more fissile material is produced than is used to maintainthe fission chain, the reactor is called a breeder

(also called forgiving). The term fault-/error-tolerant, also calledforgiving, describes the degree to which equipment faults/humaninaction (or erroneous action) can be tolerated

Fast Breeder Reactor (see Fast reactor)

Final Design Approval

Federation of Electric Power Companies

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Appendix 1:5

FFTF

Fissile

Fission products

Fissionable

FOAKE

Foolproof

Fossil fuel

FP

FPSE

FSAR

Fuel cycle

Fuel, enriched

Fuel, nuclear

Fuel reprocessing

Fast Flux Test Facility

i) of a nuclide: capable of undergoing fission by interaction withslow neutrons;

ii) of a material: containing one or more fissile nuclides

Nuclides produced either by fission or by the subsequent radioactivedecay of the nuclides thus formed

i) of a nuclide: capable of undergoing fission by any process;

ii) of a material: containing one or more fissionable nuclides

First-Of-A-Kind-Enginecring

Safe against human error or misguided human action

A term applied to coal, oil and natural gas

Fission Product(s)

Free Piston Stirling Engine

Final Safety Analysis Report

The sequence of processing, manufacturing, and transportation stepsinvolved in producing fuel for a nuclear reactor, and in processing fueldischarged from the reactor

Nuclear fuel containing uranium which has been enriched in one or moreof its fissile isotopes or to which chemically different fissile nuclides havebeen added

Material containing fissile nuclides which when placed in a reactorenables a self-sustaining nuclear chain to be achieved

The chemical or metallurgical treatment of spent (used) reactor fuel torecover the unused fissionable material, separating it from radioactivewaste. The fuel elements are chopped up and chemically dissolved.Plutonium and uranium and possibly other fissionable elements are thenseparated for further use

GA

GCR

GE

GEO

GHR

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General Atomic

Gas-Cooled Reactor

General Electric

Geostationary Earth Orbit, Geosynchronous Orbit

Gas-Cooled Heating Reactor


1995-02-13

Appendix 1:6

GIS

GLFC

GmbH

Grace period

GSOC

GWe

Graphite Impact Shell

Graphite Lunar Module Fuel Cask

German term signifying "Limited"

The period of time which a safety function is ensured without thenecessity of personnel action in the event of an incident/accident

Geosynchronous Orbit Complex

Gigawatt (109 watts) electric

H

Half-life

Heavy water

HEO

HEU

HR

HSBWR

HTR or HTGR

HTTR

HWR

HX

Hydropress EDO

The period of time required for the radioactivity of a substance to dropto half its original value; the time that it takes for half of the atoms of aradioactive substance to decay. Measured half-lives vary from millionthsof a second to billions of years

Deuterium oxide (D2O): water containing significantly more than thenatural proportion (1 in 6500) of heavy hydrogen (deuterium) atoms toordinary hydrogen atoms

High Earth Orbit

Highly-Enriched Uranium

Heat Rejection

Hitachi Small Boiling Water Reactor

High-Temperature Gas-Cooled Reactor. A graphite-moderated, helium-cooled advanced reactor that utilizes low enriched uranium

High Temperature Test Reactor

Heavy-Water Reactor. Heavy water is used as a moderator in certainreactors because it slows down neutrons effectively and also has a lowcross section for absorption of neutrons

Heat Exchange

Reactor Plant Design Bureau, Podolsk, Russia

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Appendix 1:7

I&C

IAEA

ICRP

IEA

IFR

msHASA

INET

Inherent safetycharacteristics

INSRP

Isotopes

IUS

Instrumentation & Control system

International Atomic Energy Agency

International Commission on Radiological Protection

International Energy Agency

Integral Fast Reactor

Intermediate Heat Exchange

International Institute for Applied System Analysis

Institute Nuclear Energy Technology, China

Safety achieved by die elimination of a specified hazard by means of thechoice of material and design concept

Interagency Nuclear Safety Review Panel

Nuclides having the same atomic number (i.e. identical chemicalelement) but different mass numbers; e.g. 92-Uranium-235 and92-Uranium-238. Isotopes have the same chemical properties butslightly different physical properties

Inertia] Upper Stage

JJAERI Japan Atomic Energy Research Institute

K

KEPCO

keV

kT

kW

kWh

Kansai Electric Power Company

Thousands electron volts

Kiloton

Kilowatt

Kilowatt hour (103 watt hour)

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Appendix 1:8

LANL

LCP

LEO

LET

LEU

LH2

LiH

LINAC

Linear accelerator

LLFP

LMCPBR

LMFBR

Load Factor

LOCA

LOF

LOHS

LOT

LRC

LWGR

LWR

Los Alamos National Laboratory

Large Communications Platform

Low Earth Orbit

Linear Energy Transfer

Low Enriched Uranium

Liquid hydrogen

Lithium hydrogen

Linear Accelerator

A long straight tube (or series of tubes) in which charged particles(ordinarily electrons or protons) gain in energy by action of oscillatingelectromagnetic fields

Long Lived Fission Products

Liquid Metal Cooled Pebble Bed Reactor

Liquid Metal Fast Breeder Reactor. A fast reactor that employs liquidmetal (sodium) as a coolant The sodium in the primary loop passesthrough the reactor and transfers its heat to sodium in a secondary loop.This sodium then heats water in a tertiary loop which produces steamand drives a turbine. See also Fast Reactor

The load factor of a nuclear unit or station for a given period of time isthe ratio of the energy that is produced during the period considered tothe energy that it could have produced at maximum capacity undercontinuous operation during the whole of that period. Also calledCapacity Factor

Loss Of Coolant Accident

Loss Of Flow

Loss Of Heat Sink

Lunar Orbital Terminal

Lunar Resources Complex

Light Water Graphite Reactor. A reactor that uses ordirary water as acoolant and graphite as a modertor and utilizes slightly enrichedUranium-235 (e.g. Russian multi-channel RBMK reactor plant)

Light-Water Reactor. A nuclear reactor that uses ordinary water as botha moderator and a coolant and utilizes slightly enriched Uranium-235fuel. There are two commercial LWR types: the boiling-water reactor(BWR) and the pressurized water-reactor (PWR)

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1995-02-13

Appendix 1:9

M

MAP

MARS

MCP

MeV

MHTGR

MHW

Minor actinides

Moderator

MOX fuel element

MPD

MSBR

MSPWR

MSR

MW

MWe

MWh

MWt

Minimum Attention Plant

Multipurpose Advanced Reactor Inherently Safe

Main Circulation Pump

Million electron volts

Modular HTGR, see HTR

Multihundred watt

The transuranic elements minus plutonium. Usually this term is used torefer to neptunium, americium, and curium. Some also refer to these asthe "minor" transuranics. Plutonium is the dominant transuranic, butthese minor transuranics contribute comparable radioactivity in spentfuel

A material, such as ordinary water, heavy water, beryllium, graphite andsome others used in a nuclear reactor to slow down fast neutrons sofissile nuclei can more easily and efficiently capture them, thus inaeasingthe likelihood of further fission

Mixed Oxide fuel element Fuel element in which fuel is an intimatemixture of uranium and plutonium oxides

Maximum Permissible Dose

Molten Salt Breeder Reactor, see MSR

Mitsubishi Simplified PWR

Molten Salt Reactor. A nuclear reactor that uses fluid fuel, e.g. moltensalt fluorides of Li, Be, Th, U.

Megawatt

Megawatt (106 watts) electric

Megawatt hour (106 watts hour)

Megawatt (106 watts) thermal

N

NaK

NASA

NC

NCV

Sodium-potassium eutectic mixture

(U.S.) National Aeronautics and Space Administration

Natural Circulation

Nozzle Control Valve

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1995-02-13

Appendix 1:10

NEA

NEP

NEPSTP

NERVA

Neutrons, fast

Neutrons, slow

Neutrons, thermal

NF

NHP

NHPD

NPP

NPPS

NRC

NRDS

NSSS

NTS

Nuclear energy

Nuclear fuel

Nuclear reaction

Nuclear reactor

Nuclide

OECD Nuclear Energy Agency

Nuclear Electric Propulsion

Nuclear Electric Propulsion Space Test Program

Nuclear Engine for Rocket Vehicle Application

Neutrons of kinetic energy greater than some specified value. This valuemay vary over a wide range and will be dependent upon the application,such as reactor physics, shielding, or dosimetry. In reactor physics thevalue is frequently chosen to be 100 000 eV (electron-Volt)

Neutrons of kinetic energy less than some specified value (see neutrons,fast). In reactor physics the value is frequently chosen to be 1 eV

Neutrons in thermal equilibrium with the medium in which they exist

Nuclear Furnace

Nuclear Heating Plant

Nuclear Heat and Power Plant

Nuclear Power Plant. A reactor or reactors together with all structures,systems and components necessary for the production of power (i.e. heator electricity)

Nuclear Power Propulsion System

(Nuclear Regulatory Commission): US body regulating the use ofnuclear energy

Nuclear Rocket Development Station

Nuclear Steam Supply System

Nevada Test Site

Energy released in nuclear reactions or transitions

See Fuel, nuclear

A reaction involving a change in an atomic nucleus, such as fission,fusion, neutron capture, or radioactive decay, as distinct from a chemicalreaction, which is limited to changes in the electron structuresurrounding the nucleus

A device in which a fission chain reaction can be initiated, maintained,and controlled. Its essential component is a core containing fissionablefuel. It is sometimes called an atomic "furnace"; it is the basic machine ofnuclear energy

Any species of atom that exists for a measurable length of time. Theterm is used synonymously with isotope. A radionuclide is a radioactivenuclide.

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Appendix 1:11

O

OECD

OKBM

OMEGA

OTV

Organisation for Economic Co-Operation and Development

Reactor Plant Design Bureau, Nizni-Novgorod, Russia

Options for Making Extra Gains from Actinides and Fission Products

Orbital Transfer Vehicle

Passive component

Passive system

PBIS

PbTe

PC

PCCS

PCI

PCIV

PCRV

PCS

PCV

Peak load

Pebble bed reactor

Per capita

PHRS

PIUS

Plutonium

PPO

A component which does not need any external input to operate

Either a system which is composed entirely of passive components andstructures or a system which uses active components in a very limitedway to initiate subsequent passive operation

Pressure Balanced Injection System

Lead telluride

Power Convenor

Passive Containment Cooling System

Pellet Clad Interaction

Prestressed Cast Iron Vessel

Prestressed Concrete Reactor Vessel

Power Conversion System

Pressure Containment Vessel

The maximum load produced by a unit or group of units in a statedperiod of time. A station used for peak load generation is a station that isnormally operated to provide power during maximum load periods only

A reactor, which utilizes spherical fuel elements

Per unit of a population

Passive Heat Removal System

Process Inherent Ultimate Safety Reactor

A heavy, radioactive, man-made metallic element with the atomicnumber 94, created by absorption of neutrons in uranium-238. Its mostimportant isotope is plutonium-239, which is fissile

Pressed Plutonium (-238) Oxide

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Appendix 1:12

Primary energy

Proton

PSA

PSI

PSM

PSOV

PUC

PUREX process

Pyroprocessing

PWR

The energy content of fuels before they are processed and converted intoforms used by the consumer. Primary energy refers to energy in the formof natural resources: water flowing over a dam, freshly mined coal,crude oil, natural gas, natural uranium. Only rarely can primary energybe used to supply final energy; one of the few forms of primary energythat can be used as final energy is natural gas

A particle with a single positive unit of electrical charge and a mass thatis approximately 1.840 times that of the electron. It is the nucleus of thehydrogen atom and a constituent of all atomic nuclei.

Probabilistic Safety Analysis

Paul Scherrer Institute, Switzerland

Power System Module

Propellant Shutoff Valve

Public Utility Commission

The plutonium and uranium extraction (PUREX) process is an aqueousprocess used in several foreign commercial and U.S. defense programsfor separating out elements in spent nuclear fuel

Nonaqueous processing carried out at high temperatures. An example ofthis is the relatively new technology being developed for reprocessing

Pressurized-Water Reactor. A light-water moderator and cooled reactorthat employs an indirect cycle; the cooling water that passes through thereactor is kept under high pressure to keep it from boiling, but it heatswater in a secondary loop that produces steam that drives the turbine

R

R-MAD

rad

Radioactive

Radioactive decay

Reactor Maintenance, Assembly and Disassembly

Radiation absorbed dose

Referring to the spontaneous transformation of one atomic nucleus intoa different nucleus or into different energy states of the same nucleus

The spontaneous transformation of one atom into a different atom orinto a different energy state of the same atom. The process results in adecrease, with time, of the original number of radioactive atoms in asample

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Appendix 1:13

Radioactive waste

Radioisotope

RBE

RBMK

RBR

RCC

RCP

RCS

RDIPE

Reactor core

Recycling

rem

Reprocessing, fuel

RHR

RPV

RRA

RRC-KI

RTG

RVACS

The unwanted radioactive materials formed by fission and other nuclearprocesses in a reactor or obtained in the processing or handling ofradioactive materials. Most nuclear waste is initially in the form of spentfuel. If this material is reprocessed, new categories or waste result: high-level, transuranic, and low-level wastes (as well as others)

A radioactive isotope. An unstable isotope of an element that decaysspontaneously, emitting radiation. Radioisotopes contained in the spentfuel resulting from the production of nuclear power generally fall intotwo categories: fission products and transuranic elements (known astransuranics, actinides, or TRU), and activation products produced byneutron absorption in structural material in the spent fuel

Relative Biological Effectiveness

See LWGR

Rotating Bed Reactor

Reinforced Concrete Containment

Reactor Coolant Pump

Reactor Coolant System

Research and Development Institute of Power Engineering, Moscow,Russia

The central portion of a nuclear reactor containing the fuel elements andthe control rods (and part of the coolant and moderator), where most ofthe energy is produced

The reuse of fissionable material, after it has been recovered by chemicalprocessing from spent reactor fuel

Roentgen equivalent man

A generic term for the chemical and mechanical processes applied to fuelelements discharged from a nuclear reactor; the purpose is to removefission products and recover fissile (uranium-233, uranium-235,plutonium-239), fertile (thorium-232, uranium-238) and other valuablematerial

Residual Heat Removal

Reactor Pressure Vessel

Rolls-Royce and Associates

Russian Research Centre "Kurchatov Institute", Moscow, Russia

Radioisotopc Thermoelectric Generator

Reactor Vessel Auxiliary Cooling System

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Appendix 1:14

SÄDS

Safeguards

SAFR

SBWR

SDR

SECURE

SES

SCLWR

SG

SiGi

Simplified design

SIR

SIS

SMF

SMR

SMS

SNAP

SNPS

SNR

SPAR

Spent fuel

SPR

SPS

STA

Subcritical Accelerator-Driven System

Term used to refer to the total set of international verifications,observations, etc, which together constitute a determination that nuclearmaterials (or, in some international agreements, facilities or othermaterials) have not been diverted from nuclear power programmes tothe production of nuclear weapons

Sodium Advanced Fast Reactor

Simplified or Small BWR. See BWR

Slowpoke Demonstration Reactor, see SES

Safe Environmentally Clean Urban Reactor

Slowpoke Energy System

Steam Cooled Light Water Reactor

Steam Generator

Silicon germanium

A system designed with a minimum number of components to achievethe related safety function and relying as little as possible on supportsystems

Safe Integral Reactor

Safety Injection System

Space Manufacturing Facility

Small and Medium-sized Reactors

Ship Mobile System

Space Nuclear Auxiliary Power program, US

Space Nuclear Power Source

Space Nuclear Reactor

Space Power Advanced Reactor

Nuclear reactor fuel that has been irradiated (used) to the extent that itcan no longer effectively sustain a chain reaction and therefore has beenremoved from the reactor for disposal. This radiation fuel containsfission products, uranium, and transuranic isotopes

Advanced Space Nuclear Power Program

Satellite Power System

Science and Technology Agency, Japan

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Appendix 1:15

STAR-C

STS

Subcritical

Space Thermionic Advanced Reactor Compact, US

Space Transportation System (Space Shuttle)

Not capable of sustaining a nuclear chain reaction, but involving somedegree of multiplication of neutrons

T - l l l

TAGS

Target

TBCV

TCT

TE

TFE

TI

TITR

TOP

TOSBWR

TPC

TPS

Transmutation

Transuranic

TRIGA

TRU

TRUEX

TSCV

Tantalum alloy (Ta-8W-2Hf)

Silver antimony germanium telluride

Material subjected to particle bombardment (as in an accelerator) inorder to induce a nuclear reaction

Turbine Bypass Control Valve

Thermionic Critical Technology program, US, see SNP

Thermoelectric

Thermionic Fuel Element

Thermionic

Thermionic Test Reactor

Transient Over Power

Toshiba Simplified Boiling Water Reactor

Transmutation Plant System

TRIGA Power Safety System, see DNHPP

The transformation (change) of one element into another by a nuclearreaction or series of reactions

An element above uranium in the Periodic Table of elements - that is,one that has an atomic number greater than 92. All transuranics areproduced artificially (during a man-made nuclear reaction) and areradioactive. They are neptunium, plutonium, americium, curium,berkelium, californium, einsteinium, fermium, mendelevium, nobelium,and lawrencium

Training of personnel, nuclear Research, Isotope production, GeneralAtomic

Transuranium elements

A chemical solvent process under development to extract transuranicsfrom high-level waste

Turbine Shutoff Control Valve

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1995-02-13

Appendix 1:16

TVA

TWh

TWS

TZM

Turbine Valve Actuator

Terrawatt (1012 watt) hour

Transmutation Waste System

Molybdenum alloy (Mo-0.6, Ti-0.1, Zr-0.035)

uU-ZrH

UO2

Uranium

USR

Uranium zirconium hydride

Uranium dioxide

A radioactive element of atomic number 92. Naturally occurring uraniumis a mixture of 99.28 per cent uranium-238,0.715 per cent uranium-235,and 0.0058 per cent uranium-234. Uranium-235 is a fissile material andis the primary fuel of nuclear reactors. When bombarded with slow orfast neutrons, it will undergo fission. Uranium-238 is a fertile materialthat is transmuted to plutonium-239 upon the absorption of a neutron.

Ultimate Safety Reactor, see MSR

V

VVER Russian version of PWR plant

wwWANO

Waste separation

WEC

WOCA

WRI

Westinghouse

World Association of Nuclear Operations

The dividing of waste into constituents by type (for example, high-level,medium-level, low-level) and/or by isotope (for example, separating outplutonium and uranium). The waste may be separated by a chemicalsolvent process such as PURES or by any of a number of other chemicalor physical processes

World Energy Council

World Outside CPE Areas (MEM ou Monde å Economic de Marché)

World Resource Institute

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Appendix 1:17

X

XE Experimental Engine

zZrH Zirconium hydride

References

Small and medium reactorsI. Status and prospectsReport OECD, NEA, 1991, Paris, France.

Safety related terms for advanced nuclear plantsReport IAEA-TECDOC-626,1991, September.

Developing Technology to Reduce Radioactive Waste May Take Decades and BeCostlyReportGAO/RCED-94-16,1993, December.

Space nuclear power systems 1989Orbit book corrroany, Inc, 1992, Malabar, Fl, US.

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