Advances in Small Modular Reactor
Technology Developments
A Supplement to:
IAEA Advanced Reactors Information System (ARIS)
2022 Edition

ADVANCES IN SMALL MODULAR REACTORTECHNOLOGY DEVELOPMENTS
2022 Edition
A Supplement to:
IAEA Advanced Reactors Information System (ARIS)
http://aris.iaea.org

DISCLAIMER
This is not an official International Atomic Energy Agency (IAEA) publication. The material has notundergone an official review by the IAEA. The views expressed do not necessarily reflect those of the
IAEA or its Member States and remain the responsibility of the contributors.
Although great care has been taken to maintain the accuracy of information contained in this publication,neither the IAEA nor its Member States assume any responsibility for consequences which may arise from its
use.
The use of particular designations of countries or territories does not imply any judgement by the publisher,
the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of thedelimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does not
imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or
recommendation on the part of the IAEA.

Printed by the IAEA in Austria
September 2022

FOREWORD
The IAEA Department of Nuclear Energy continues to facilitate efforts of Member States in the development
and deployment of small modular reactors (SMRs), recognizing their potential as a viable solution to meetenergy supply security, both in newcomer and expanding countries interested in SMRs. In this regard, balanced
and objective information to all Member States on technology status and development trends for advanced
reactor lines and their applications are collected, assessed and provided through publication of status reports
and other technical documents.
Member States, both those launching their nuclear power programme and those with an existing nuclear power
programme, keep expressing their interest in information about advances in SMR design andtechnology developments, as well as global trend of their deployment. The IAEA's Division of Nuclear
Power, which has been facilitating Member States in addressing common technologies and issues for SMRs,
plays a prominent role in convening international scientific forums and technical cooperation inthis field for the interested Member States. The activities on SMRs are further supported by specificactivities on advanced water-cooled, liquid metal-cooled fast neutron spectrum, molten salt, andhigh temperature gas-cooled reactors technology developments, as well as their non-electric applications.
The driving forces in the development of SMRs are their specific characteristics. They can be deployedincrementally to closely match increasing energy demand resulting in a moderate financial commitment for
countries or regions with smaller electricity grids. SMRs show the prospect of significant cost
reduction through modularization and factory construction which should further improve the construction
schedule and reduce costs. In the area of wider applicability, SMR designs and sizes are better suitedfor partial or dedicated use in non-electric applications such as district heating, providing heat forindustrial processes, hydrogen production or sea-water desalination. Process heat or cogeneration
results in significantly improved thermal efficiencies leading to a better return on investment. Some SMR
designs may also serve niche markets, for example by deploying microreactors to replace diesel
generators in small islands or remote regions.
Booklets on the status of SMR technology developments have been published biennially since 2012.
The objective is to provide Member States with a concise overview of the latest status of SMR designs.
This booklet is reporting the advances in design and technology developments of SMRs of all the
major technology lines within the category of SMRs. It covers land-based and marine-based watercooled reactors, high temperature gas-cooled reactors, liquid metal-cooled fast neutron spectrum reactors,
molten salt reactors, and a sub-category called microreactors with electrical power typically up to 10
MW(e). For the first time also that the booklet provides some insights on the economic challenges in
deployment of SMRs, a summary on enabling design features to facilitate SMRs' decommissioning and asummary on experimental testing for design verification and validation. The brief design description of
SMRs is provided by the responsible institute or organization and is reproduced, with permission,in this booklet.
This booklet is intended as a supplement to the IAEA Advanced Reactor Information System (ARIS),
which can be accessed at http://aris.iaea.org. Previous editions of this booklet published in support of ARIS
are listed in Annex X.
This publication was developed by Nuclear Power Technology Development Section, Division of
Nuclear Power of the IAEA's Department of Nuclear Energy in cooperation with Member States.
The IAEA officers responsible for this publication were Y. Zou and M.H. Subki of the Division of
Nuclear Power.

TABLE OF CONTENTSINTRODUCTION.................................................................................................................... 1
PART I.1. WATER COOLED SMALL MODULAR REACTORS (LAND BASED) ............ 9
CAREM (CNEA, Argentina) ............................................................................................................ 11
ACP100 (CNNC,China) ................................................................................................................... 15
CANDU SMRTM (Candu Energy Inc, Canada) ................................................................................. 19
CAP200 (SPIC/SNERDI, China)...................................................................................................... 23
DHR400 (CNNC, China) .................................................................................................................. 27
HAPPY200 (SPIC, China) ................................................................................................................ 31
NHR200-II (Tsinghua University and CGN, China) ....................................................................... 35
TEPLATORTM (UWB Pilsen & CIIRC CTU, Czech Republic) ....................................................... 39
NUWARDTM (EDF, France) ............................................................................................................. 43
IMR (Mitsubishi Heavy Industries, Japan) ....................................................................................... 47
i-SMR (KHNP & KAERI, Republic of Korea)................................................................................. 51
SMART (KAERI, Republic of Korea and K.A.CARE, Saudi Arabia) ............................................. 55
RITM-200N ( JSC “Afrikantov OKBM”, Russian Federation)........................................................ 59
VK-300 (NIKIET, Russian Federation) ............................................................................................ 63
KARAT-45 (NIKIET, Russian Federation) ....................................................................................... 67
KARAT-100 (NIKIET, Russian Federation) ..................................................................................... 71
RUTA-70 (NIKIET, Russian Federation) ......................................................................................... 75
STAR (STAR ENERGY SA, Switzerland) ....................................................................................... 79
Rolls-Royce SMR (Rolls-Royce SMR Ltd, United Kingdom)......................................................... 83
VOYGRTM (NuScale Power Corporation, United States of America) .............................................. 87
BWRX-300 (GE-Hitachi Nuclear Energy, USA and Hitachi-GE Nuclear Energy, Japan) ............... 91
SMR-160 (Holtec International, United States of America) ............................................................. 95
Westinghouse SMR (Westinghouse Electric Company LLC, United States of America) ................ 99
mPower (BWX Technologies, Inc., United States of America) ..................................................... 103
OPEN20 (Last Energy Inc., United States of America) .................................................................. 107

PART I.2. WATER COOLED SMALL MODULAR REACTORS (MARINE BASED) ... 111
KLT-40S (JSC “Afrikantov OKBM”, Russian Federation) ............................................................ 113
ACPR50S (CGNPC, China) ........................................................................................................... 117
ACP100S (CNNC, China) .............................................................................................................. 121
BANDI-60 (KEPCO E&C, Republic of Korea) ............................................................................. 125
ABV-6E (JSC “Afrikantov OKBM”, Russian Federation) ............................................................ 129
RITM-200M (JSC “Afrikantov OKBM”, Russian Federation) ...................................................... 133
VBER-300 (JSC “Afrikantov OKBM”, Russian Federation) ......................................................... 137
SHELF-M (NIKIET, Russian Federation) ...................................................................................... 141

PART II. HIGH TEMPERATURE GAS COOLED SMALL MODULAR REACTORS .... 145

HTR-PM (Tsinghua University, China) .......................................................................................... 147
STARCORE (StarCore Nuclear, Canada, United Kingdom and United States of America) .......... 151
JIMMY (JIMMY ENERGY SAS, France) ..................................................................................... 155
GTHTR300 (JAEA Consortium, Japan) ......................................................................................... 159
GT-MHR (JSC “Afrikantov OKBM”, Russian Federation) ........................................................... 163
MHR-T Reactor (JSC “Afrikantov OKBM”, Russian Federation)................................................. 167
MHR-100 (JSC “Afrikantov OKBM”, Russian Federation) .......................................................... 171
AHTR-100 (Eskom Holdings SOC Ltd., South Africa) ................................................................. 177
PBMR®-400 (PBMR SOC Ltd., South Africa) ............................................................................... 181
HTMR100 (STL Nuclear (Pty) Ltd., South Africa) ........................................................................ 185
EM2 (General Atomics, United States of America) ........................................................................ 189
FMR (General Atomics, United States of America) ....................................................................... 193
Xe-100 (X-energy, LLC, United States of America) ...................................................................... 197
SC-HTGR (Framatome Inc., United States of America) ................................................................ 201
PeLUIt/RDE (BRIN, Indonesia) ..................................................................................................... 205
HTR-10 (Tsinghua University, China) ............................................................................................ 209
HTTR (JAEA, Japan)...................................................................................................................... 213

PART III. LIQUID METAL COOLED FAST NEUTRON SPECTRUM SMALL MODULAR
REACTORS ........................................................................................................................... 217
BREST-OD-300 (NIKIET, Russian Federation) ............................................................................ 219
ARC-100 (ARC Clean Energy, Canada)......................................................................................... 223
4S (Toshiba Energy Systems & Solutions Corporation, Japan) ...................................................... 227
MicroURANUS (UNIST, Republic of Korea) ................................................................................ 231
LFR-AS-200 (newcleo srl, Italy) .................................................................................................... 235
SVBR (JSC AKME Engineering, Russian Federation) ................................................................. 239
SEALER-55 (LeadCold, Sweden) .................................................................................................. 243
Westinghouse Lead Fast Reactor (Westinghouse Electric Company, USA) ................................... 247

PART IV. MOLTEN SALT SMALL MODULAR REACTORS .......................................... 251
IMSR400 (Terrestrial Energy Inc., Canada) ................................................................................... 253
SSR-W (Moltex Energy, Canada) ................................................................................................... 257
smTMSR-400 (SINAP, CAS, China).............................................................................................. 261
CMSR (Seaborg Technologies, Denmark) ...................................................................................... 265
Copenhagen Atomics Waste Burner (Copenhagen Atomics, Denmark) ......................................... 269
FUJI (International Thorium Molten-Salt Forum, Japan) ............................................................... 273
THORIZON (THORIZON B.V., Netherlands) ............................................................................... 277
SSR-U (Moltex Energy, United Kingdom) ..................................................................................... 281
KP-FHR (Kairos Power, United States of America) ....................................................................... 285
Mk1 PB-FHR (UC Berkeley, United States of America) .............................................................. 289
MCSFR (Elysium Industries, USA)................................................................................................ 293

LFTR (Flibe Energy, United States of America) ............................................................................ 297
ThorCon (ThorCon International, United States of America and Indonesia) ................................. 301

PART V. MICROREACTORS .............................................................................................. 305
Energy Well (Centrum výzkumu Řež s.r.o., Czech Republic) ........................................................ 307
MoveluX (Toshiba Energy Systems & Solutions Corporation, Japan) .......................................... 311
ELENA (NRC “Kurchatov Institute”, Russian Federation)........................................................... 315
UNITHERM (NIKIET, Russian Federation) ................................................................................. 319
AMR (STL Nuclear (Pty) Ltd, South Africa) ................................................................................ 323
LFR-TL-30 (newcleo Ltd., United Kingdom) ................................................................................ 327
U-Battery (Urenco, United Kingdom) ............................................................................................ 331
Aurora Powerhouse Product Line (Oklo Inc., United States of America) ...................................... 335
HOLOS-QUAD (HolosGen LLC, United States of America) ........................................................ 337
MARVEL Research Microreactor (Idaho National Laboratory, United States of America)........... 341
MMR® (Ultra Safe Nuclear Corporation, United States of America) ............................................. 345
Westinghouse eVinci™ Micro Reactor (Westinghouse Electric Company LLC, USA) .................. 349
ANNEX I. Summary of Global SMR Technology Development and Deployment ........................... 353
ANNEX II. Power Range of SMR Designs ........................................................................................ 358
ANNEX III. Comparison of Main Technical Characteristics among several SMR Designs .............. 364
ANNEX IV. Non-Electric Applications using Small Modular Reactors ............................................ 370
ANNEX V. Economic Challenges in Deployment of Small Modular Reactors ................................. 374
ANNEX VI. Fuel Cycle Approach adopted in SMRs ......................................................................... 378
ANNEX VII. Spent Fuel, Waste Management and Disposal Plans adopted for SMRs ...................... 385
ANNEX VIII. Enabling Design Features to Facilitate SMR Decommissioning ................................ 390
ANNEX IX. Experimental Testing for Design Verification and Validation........................................ 394
ANNEX X. Bibliography.................................................................................................................... 401
ANNEX XI. Acronyms ....................................................................................................................... 405
ANNEX XII. Acknowledgement ........................................................................................................ 411

INTRODUCTION
The IAEA’s Department of Nuclear Energy’s Section for Nuclear Power Technology Development is tasked
to facilitate efforts of Member States in identifying key enabling technologies for the development of advanced
reactor lines and addressing their key challenges for near-term deployment. By establishing international
networks and ensuring the coordination of the Member States’ experts, publications on international
recommendations and guidance focusing on needs of the Member States are issued.
Climate change presents us with a stark challenge, that is to reduce greenhouse gas emissions much fasterthan has been done so far or face the increasingly catastrophic consequences of an inexorably warming
planet. The world needs to harness all low-carbon sources of energy to meet the Paris Agreement goal of
limiting the rise in global temperatures to well below 2°C above pre-industrial levels. Use of renewables
such as wind and solar power is growing, while nuclear power continues to make a significant contribution
to energy supply, energy security and grid stability. Of all the low-carbon energy sources, nuclear power is
one of the few, if not the only one, that can generate at scale all the main energy carriers: electricity, heat,and hydrogen. Advanced nuclear reactor technologies, particularly Small Modular Reactors (SMRs)
including microreactors, are designed to not only produce base load and dispatchable carbon-freeelectricity but also to supply other clean energy products needed to decarbonize energy-intensive sectors
such as the transport sector, the building sector, industrial heat applications, and desalination. SMRs arealso designed with enhanced operability to enable flexible operation with variable renewables.
There is increasing interest in SMRs and their applications. During the International Conference on Climate
Change and the Role of Nuclear Power held in September 2019, SMRs were considered by many
Member States as a potential viable nuclear option to contribute in mitigating the climate change. Following
on this conference, in April 2021, the IAEA launched the IAEA Platform on Small Modular Reactors andtheir Applications (the SMR Platform), a mechanism that coordinates the IAEA’s activities in this field andprovides a ‘one‑stop shop’ for Member States and other stakeholders. The SMR Platform offers expertisefrom the entire Agency, encompassing all aspects relevant to the development, early deployment, andoversight of SMRs and their applications.
SMRs are advanced reactors with a power capacity of typically up to 300 MW(e) per unit, which is about
one-third of the generating capacity of traditional nuclear power reactors and whose components andsystems can be shop fabricated and then transported as modules to the sites for installation as demand arises.
Most of the SMR designs adopt advanced safety features and are deployable either as a single or multimodule plant. SMRs are under development for all principal reactor technology lines: water-cooled reactors,
high temperature gas-cooled reactors, liquid metal-cooled fast neutron spectrum reactors, molten salt
reactors, and microreactors. The key driving forces of SMR development are fulfilling the need forflexible power generation for a wider range of users and applications, replacing ageing fossil-fired units,enhancing safety performance, and offering better economic affordability.
SMRs are envisioned for niche electricity or energy markets where large reactors would not be viable. SMRs
could provide cogeneration for heat, hydrogen production, desalination for small electricity grids,
remote and off-grid areas, and enabling hybrid nuclear-renewables energy systems. Through modularisation
technology, SMRs target at the economies of serial production with shorter installation or construction time.
Near-term deployable SMRs are expected to have safety performance comparable or better to thatof evolutionary reactor designs.
With smaller footprints, SMRs are expected to have flexibility in siting and allow them to be tailored to theenergy needs of regional or industrial clusters. The modularity and advanced safety features make SMRs
attractive to industries and countries with smaller grid sizes and little to no expertise of operating nuclear
plants. Some transportable turnkey systems are being developed and deployed that are completely built in a
shipyard factory, delivered to remote sites, or exported to other countries as a marine plant for plug and playfor electricity and heat supply. The possibility of siting SMRs in remote, off-grid communities enhance theaccess of such community to clean electricity and heat. These advantages relate well with the floating power
plants concepts based on SMRs.
Though significant advancements have been made in various SMR technologies in recent years, sometechnical issues still attract considerable attention in the industry. These include control room staffing and1human factor engineering for multi-module SMR plants, applicability of existing codes and standards,
manufacturing approach for novel components, and back-end solutions for fuel cycle. Somepotential deployment advantages of SMRs such as reduced size of Emergency Planning Zone (EPZ) or
single operator monitoring several modules are under discussion by nuclear regulators. Although SMRs aredesigned for lower upfront capital cost per unit, their economic competitiveness is still to be proven.
There are more than eighty (80) SMR designs under development and deployment at different stages in 18
Member States. Several major milestones have been reached in SMR technology deployment. The
Akademik Lomonosov floating power unit in the Russian Federation with two-module KLT-40S wasconnected to the grid in December 2019 and started commercial operation in May 2020. The HTR-PM
demonstrator in China was connected to the grid in December 2021 and is expected to reach fullpower operation by the end of 2022. The CAREM25 in Argentina is under construction and is
expected to reach first criticality in 2026. The construction of ACP100 in China started in July 2021 andis targeted to start commercial operation by the end of 2026. The construction of BREST-OD-300 in
Russian Federation began in June 2021 and is planned to be completed in 2026. The NuScale Power
Module™ in the United States has received Standard Design Approval from U.S. NRC in September 2020.
The NRC has directed to issue a final rule that certifies NuScale’s SMR design for use in the United States.
This booklet provides a brief introductory information and technical descriptions of various SMR designs
and technologies under different stages of development and deployment. To assist the readers inbetter understanding the status of development and deployment, Table 1 lists all the eighty-three (83)designs included in this booklet, along with the output capacity, reactor type, designer organisation, countryof origin and status of development or deployment.
The 2022 edition comprises five (5) parts. Parts I to IV are arranged in the order of the different typesof coolants, with Part I divided into two sub-parts (land based and marine based). Part V is dedicatedto microreactors, a sub-category of SMR which either has a relatively smaller output capacity (typically lessthan 10 MWe) or has specific applications, such as powering remote regions. The part of microreactors
includes technologies that use different coolants and/or fuel design, such as heat pipe.
Table 1 Design and Status of SMRs included in this Booklet
OutputMW(e)DesignType

Designer

Country

StatusPART I.1: WATER COOLED SMALL MODULAR REACTORS (LAND BASED)
CAREM

30

ACP100125

CANDU SMRTM
CAP200
DHR400300> 200
400 MW(t)HAPPY200
NHR200-II

200 MW(t)200 MW(t)TEPLATORTMNUWARDTM

2

< 150
MW(t)
2 × 170

IMR

350

i-SMR

170

SMART

107

RITM-200N

55

VK-300KARAT-45
KARAT-100

250
45 – 50100

Integral
PWR
Integral
PWR
PHWR
PWR
PWR
(pool type)
PWR
Integral
PWR
HWR
Integral
PWR
PWR
Integral
PWR
Integral
PWR
Integral
PWR
BWR
BWR
BWR

CNEA

Argentina

Under construction

CNNC/NPIC

China

Under construction

Candu Energy Inc.
SPIC/SNERDI
CNNC

CanadaChina
China

Conceptual Design
Basic Design
Basic Design

SPIC
Tsinghua University
and CGNUWB Pilsen & CIIRC
CTU
EDF

China
China

Detailed Design
Basic Design

Czech Republic

Conceptual Design

FranceConceptual Design

MHI

Japan

KHNP and KAERI

Republic of Korea

Conceptual Design
Completed
Conceptual design

KAERI and
K.A.CARE
JSC Afrikantov
OKBM, Rosatom
NIKIET
NIKIET
NIKIET

Republic of Korea
and Saudi Arabia
Russian Federation
Russian Federation
Russian Federation
Russian Federation

Detailed Design
Detailed Design
Completed
Detailed Design
Conceptual Design
Conceptual Design

Design
RUTA-70

OutputMW(e)70 MW(t)

STAR

10

Type

Designer

Country

StatusPWR(pool type)
LWR(pressuretube)
PWR
Integral
PWR

NIKIET

Russian Federation

Conceptual Design

STAR ENERGY SA

Switzerland

Basic design

Rolls-Royce SMR Ltd.
NuScale Power
Corporation

UKUSA

USA and Japan

Detailed Design
Equipment
Manufacturing inprogress
Detailed Design

USA

Rolls-Royce SMR
VOYGRTM470
4/6/12 × 77

BWRX-300

270 – 290

BWR

SMR-160160

PWR

GE-Hitachi
Nuclear Energy and
Hitachi-GE Nuclear
EnergyHoltec International

Westinghouse SMR> 225

mPower2 × 195

OPEN2022

Integral
PWR
Integral
PWR
PWR

Westinghouse ElectricCompany LLC
BWX Technologies,
Inc
Last Energy Inc.

USA

USA

Preliminary Design
Completed
Conceptual Design
Completed
Conceptual Design

USA

Detailed Design

PART I.2: WATER COOLED SMALL MODULAR REACTORS (MARINE BASED)
KLT-40S
ACPR50S

2 × 3550

ACP100S

125

BANDI-60
ABV-6E

60
6–9

RITM-200M

50

VBER-300

325

SHELF-M

up to 10PWR
PWR
(loop type)
Integral
PWR
PWR
PWR

Integral
PWR
Integral
PWR
Integral
PWR

JSC Afrikantov OKBM
CGNPC

Russian Federation
China

In Operation
Detailed Design

CNNC/NPIC

China

Basic Design

KEPCO E&C
JSC Afrikantov
OKBM, Rosatom
JSC Afrikantov
OKBM, Rosatom
JSC Afrikantov
OKBM, Rosatom
NIKIET

Republic of Korea
Russian Federation

Conceptual Design
Final design

Russian Federation
Russian Federation

Basic Design
Completed
Licensing Stage

Russian Federation

Basic Design

PART II: HIGH TEMPERATURE GAS COOLED SMALL MODULAR REACTORSHTR-PM

210

STARCORE

14/20/60

JIMMY
GTHTR300

10 – 20MW(t)100 – 300

GT-MHR

288

MHR-T
MHR-100
AHTR-100

4 × 205.5
25 – 87
50

PBMR-400165

HTMR100

35

EM2
FMR
Xe-100

265
50
82.5

SC-HTGR272

PeLUIt / RDE
HTR-1040 MW(t)2.5

HTGR(pebble bed)
HTGR
(prismatic)
HTGR(prismatic)
HTGR(prismatic)
HTGR(prismatic)
HTGR
HTGR
HTGR
(pebble bed)
HTGR
(pebble bed)
HTGR
(pebble bed)
GFR
GFR
HTGR
(pebble bed)
HTGR
(prismatic)
HTGR(pebble bed)
HTGR
(pebble bed)

INET, Tsinghua
University
StarCore Nuclear

China

In operation

CanadaJIMMY ENERGY SAS

France

Pre-Conceptual
DesignDetailed Design

JAEA Consortium

Japan

Basic Design

JSC Afrikantov OKBM

Russian Federation

JSC Afrikantov OKBM
JSC Afrikantov OKBM
Eskom Holdings SOC
Ltd.
PBMR SOC Ltd.

Russian Federation
Russian Federation
South Africa

STL Nuclear (Pty) Ltd.

South Africa

Preliminary Design
Completed
Conceptual Design
Conceptual Design
Conceptual Design
Completed
Preliminary Design
Completed
Basic Design

General Atomics
General Atomics
X-Energy LLC

USA
USA
USA

Conceptual Design
Conceptual Design
Basic Design

Framatome, Inc.

USA

Preliminary Design

BRIN

Indonesia

Conceptual Design

INET, Tsinghua
University

China

OperableSouth Africa

3

Design
HTTR

OutputMW(e)30 MW(t)Type

Designer

Country

StatusHTGR(prismatic)

JAEA

Japan

In operation

PART III: LIQUID METAL COOLED FAST NEUTRON SPECTRUM SMALL MODULAR REACTORS
BREST-OD-300

300

ARC-100

100

4S

10

MicroURANUS

20

LFR-AS-200
SVBR

200
100

SEALER-55
Westinghouse LFR

55
450

LMFR
(pool type)
LMFR(pool type)
LMFR(pool type)
LBE-cooled
Reactor
LMFR
LMFR
LMFR
LMFR
(pool type)

NIKIET

Russian Federation

Under Construction

ARC Clean Energy

CanadaPreliminary Design

Toshiba Energy
Systems &Solutions Corporation
UNIST

Japan

Detailed Design

Republic of Korea

Conceptual Design

Italy
Russian Federation

Conceptual Design
Detailed Design

SwedenUSA

Conceptual Design
Conceptual Design

newcleo srl
JSC AKME
Engineering
LeadCold
Westinghouse ElectricCompany, LLC.

PART IV: MOLTEN SALT SMALL MODULAR REACTORSIMSR400
SSR-W
smTMSR-400

168

MSR
MSR (static
fuelled)
MSR

CMSR

100

MSR

Seaborg Technologies
ApS

Denmark

Copenhagen Atomics
Waste Burner
FUJI

20 MW(t)
200

MSR
MSR

Copenhagen Atomics
ITMSF

DenmarkJapan

THORIZON
SSR-U
KP-FHR
Mk1 PB-FHR

40 – 120
16
140
100

MSR
MSR
FHR
FHR

THORIZON
Moltex Energy
KAIROS Power, LLC.
UC Berkeley

Netherlands
UK
USA
USA

MCSFR

50 / 200 /400 / 1200250250

MSR (fast
spectrum)
MSR
MSR

Elysium Industries

USA

Flibe Energy, Inc.
ThorCon International

USA
USA and Indonesia

Conceptual Design
Preliminary Design
Completed

LFTR
ThorCon2 × 195
300

Terrestrial Energy Inc.
Moltex Energy

CanadaCanada

Detailed Design
Conceptual Design

CAS/SINAP

China

Pre-Conceptual
DesignConceptual Design
Detailed Design
Preliminary Design
Completed
Conceptual Design
Basic Design
Conceptual Design
Pre-Conceptual
DesignConceptual Design

PART V: MICROREACTORS
Energy Well

8

FHTR

Centrum výzkumu Řež

Czech RepublicMoveluX

3–4

Heat Pipe(sodium)

Japan

ELENA

0.068

PWR

Russian Federation

Conceptual Design

UNITHERM
AMR

6.6
3

Russian Federation
South Africa

LFR-TL-30
U-Battery
AuroraHOLOS-QUAD
MARVEL

30
4
1.5 – 50100.015 –
0.027

PWR
HTGR
(prismatic)
LMFR
HTGR
LMFR
HTGR
LMFR

Toshiba Energy
Systems & Solutions
Corporation
National ResearchCentre“Kurchatov Institute”
NIKIET
STL Nuclear (Pty) Ltd.

Pre-Conceptual
DesignConceptual Design

MMRTM

>5
and > 10
2 – 3.5

Conceptual Design
Pre-conceptual
designConceptual Design
Conceptual Design
Detailed Design
Detailed Design
Equipment
manufacturing inprogress
Basic Design

WestinghouseeVinciTM

4HTGR
Heat Pipe

newcleo Ltd.
Urenco
OKLO, Inc.
HolosGen LLC
Idaho National
Laboratory

UK
UK
USA
USA
USA

Ultra Safe Nuclear
Corporation
Westinghouse ElectricCompany, LLC.

USA
USA

Conceptual Design
CompletedPart I.1: Land-based water-cooled SMRs. This part presents notable water-cooled SMR designs fromvarious configurations of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. Thesedesigns make advantage of mature technology as most of the large power plants in operation consist of watercooled reactors. There are twenty-five (25) land-based water-cooled SMR designs from twelve (12) Member
States described in this booklet. These designs include integral PWR, PHWR, compact PWR, loop-type PWR,
BWRs, and pool-type PWR for district heating. An integral PWR with natural circulation, CAREM, is underconstruction for first criticality in 2026. Another integral PWR with design simplification, ACP100, started
construction in 2021 and is targeting commercial operation in 2026. More designs are being prepared for nearterm deployment, including NuScale VOYGR and BWRX-300 in the USA, Rolls-Royce SMR in the UK and
NUWARD in France.
Part I.2: Marine-based water-cooled SMRs. This part presents concepts that can be deployed in a marine
environment, either as barge-mounted floating power unit or immersible underwater power unit. This subcategory provides many flexible deployment options. There are eight (8) marine-based water-cooled SMRs.
Some of them have been deployed in nuclear icebreaker ship. KLT-40S, deployed in the Akademik Lomonosov
floating nuclear power plant which started commercial operation in Pevek of Russian Federation in May 2020,
is from this sub-category and is the first SMR design connected to the grid.
Part II: High temperature gas-cooled SMRs: This part provides information on modular HTGRs underdevelopment and in operation. HTGRs provide high temperature heat (≥750°C) that can be utilised for more
efficient electricity generation, a variety of industrial applications as well as cogeneration. Fourteen (14)
HTGR-type SMRs are described in this booklet, including China’s HTR-PM, which was connected to grid in
December 2021 and targets at full power operation by the end of 2022. This part also includes three (3) HTGR
test reactors, two of which have already been in operation for technology testing purposes respectively in Japanand China for over twenty years.
Part III: Liquid metal-cooled fast neutron spectrum SMRs. This part presents eight (8) SMR designs that
adopt fast neutron spectrum with liquid metal coolants, including sodium, pure lead and lead-bismuth eutectic.
Tangible advances in technology development and deployment on SMRs in this category have been made. The
BREST-OD-300, a lead-cooled fast neutron reactor, is in the process of construction which is planned to becompleted in 2026 at a site in Seversk, Russian Federation. This is a demonstration prototype project for future
design with large power to enable a closed nuclear fuel cycle.
Part IV: Molten salt SMRs. This part highlights thirteen (13) SMR designs from molten salt fuelled andcooled advanced reactor technology, which is also one of the six Generation IV designs. MSRs promise many
advantages including enhanced safety due to inherent property of molten salt, low-pressure single-phase
coolant system that eliminates the need of large containment, a high temperature system that results in highefficiency, and flexible fuel cycle. Several MSR designs are conducting preliminary licensing activities or
preliminary engagement with regulators in Canada, Denmark, Netherlands, the UK and the USA.
Part V: Microreactors. This booklet carries on a dedicated part to present advances on microreactors. An
unprecedented development trend emerged on very small SMRs designed to generate electrical power oftypically up to 10 MW(e). Different types of coolant, including light water, helium, molten salt and liquid
metal, are adopted by microreactors. Heat pipe is another proposed cooling system option. Several designs areundertaking licensing activities in Canada and the USA for near-term deployment. In 2019, a site initial
application was submitted by Global First Power (GFP) to Canadian Nuclear Safety Commission (CNSC) fora single small modular reactor using USNC’s Micro Modular Reactor (MMR) technology at Chalk River.
Microreactors serve future niche electricity and heat markets, such as powering micro-grids and remote offgrid areas, restoring power quickly in communities affected by natural disasters and supporting faster
restoration of critical services (e.g., hospitals, water supply) and seawater desalination. Twelve (12)microreactor designs are included and discussed in this booklet.
For this booklet, an effort has been made to present all SMR designs within the above categories. Eachtechnical description consists of an introduction, design philosophy, target applications, a table of major design
parameters, main design features, safety features, plant safety and operational performances, instrumentationand control system, plant layout arrangement, design and licensing status, fuel cycle approach, waste
management and disposal plan, and development milestones. For the first time in this biennial booklet, many5

designs also provide information on testing conducted for design verification and validation. Even thoughsome SMR designs don’t present reportable new milestones or advances, they are still included in the booklet
for the benefit of Member States’ readers by providing lessons learned in design development. Not all reactor
designs presented in this booklet can be strictly categorised as SMRs. Some strongly rely on proventechnologies of operating large capacity reactors, while others do not apply modular design. They are presented
in this booklet by reason of completeness and foreseen niche markets.
This booklet is enriched with a set of annexes that provide readers with various figures and tables to understandthe essential technical aspects of SMR designs. The description of each annex is given below. The concepts
that have been mentioned in these annexes are just examples useful to elaborate on possible options. The
summaries provided have not undergone an official review by the IAEA. The views expressed do notnecessarily reflect those of the IAEA or its Member States but remain the responsibilities of the contributors.
Annex I provides figures that summarise global SMR technology development and deployment. It includes aworld map of SMR developers, general deployment timeline that extends to early 2030s, government andprivate sectors on SMR technology development, SMR designs across the world’s regions, and a chart that
presents stage of design or deployment of SMRs in terms of their output capacity.
Annex II provides figures showing power ranges of SMR designs of above-mentioned five (5) categories.
Annex III provides tables comparing main technical characteristics among several SMR designs of the same
category.
Annex IV recaps SMR designs for different non-electric applications. Typical reactor core exit coolant
temperature of different reactor types is illustrated alongside the temperature of various non-electric
applications.
Annex V provides a summary on the economic challenges in deployment of SMRs. It provides an overview
of the economic challenges facing SMR during the key phases of a typical project – from the project
development phase to construction and commissioning to the operation phase – with a focus on costs and cost
drivers, funding and financing issues, and economic impact.
Annex VI provides a summary on fuel cycle approaches adopted in SMR designs, furnished with a table thatcategorises different SMRs according to fuel cycle types (open or closed cycle), refuelling cycle, enrichmentlevel, spent fuel processing and conditioning, the use of Thorium-cycle and/or Plutonium disposition and theuse of spent fuel as fuel.
Annex VII provides a summary of spent fuel, waste management and disposal plans adopted for SMRs. It
contains a table on the adopted approaches, i.e., volume reduction and conditioning, waste processing, storage
approach, spent fuel pool cooling mechanism, spent fuel take back option, and market potential.
Annex VIII provides a summary on enabling design features to facilitate SMRs’ decommissioning. It
introduces the decommissioning aspect of SMR designs. It shares lessons learned and aspects related to SMR
designs. Besides, a general discussion on SMRs for decommissioning presents interconnected factors in thisregard.
Annex IX provides a summary on experimental testing for design verification and validation. The advanced
design features of SMRs of major technology lines are verified by a comprehensive technology validation
program that includes safety tests and performance tests. This annex features a table that lists experimental
testing performed for validation of some integral PWR designs.
Annex X summarises other booklets previously published in support of the IAEA’s Advanced Reactor
Information System (ARIS, http://aris.iaea.org).
Annex XI contains a list of commonly used acronyms.
Annex XII refers to the acknowledgement to the organisations which provided design information and theorganisations which helped develop this edition; and to the IAEA colleagues who provided inputs to theannexes or reviewed the contents.

6This booklet is a supplement to the IAEA’s ARIS which provides more detailed information on different SMRs
under development and deployment. The technical description and major technical parameters were provided
by the design organisations without validation or verification by the IAEA. All figures, illustrations and tables
in technical description of each design were also provided by the design organisations.
It is hoped that this booklet will be useful to the Member States with a general interest in SMRs, as well as tothe countries looking for structured technical information of one or more specific SMR designs. It should
also furtherly promote contributions to and the use of the IAEA’s ARIS.

78

PART I.1. WATER COOLEDSMALL MODULAR REACTORS(LAND BASED)

9

10

CAREM (CNEA, Argentina)
All content included in this section has been provided by, and is reproduced with permission of CNEA, Argentina

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

CNEA, Argentina

Coolant/moderator

Light water / Light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

100 / ~30 (CAREM25)

Integral PWR

Natural circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

12.25 / 4.7 (CAREM25)

Number of fuel assemblies in the core61 (CAREM25)

Fuel enrichment (%)

3.1% (CAREM25)

Core Discharge Burnup (GWd/ton)

24 (CAREM25)

Refuelling Cycle (months)

14 (CAREM25)

Reactivity control mechanism
Approach to safety systems

Control rod driving
mechanism (CRDM) only
Passive

Design life (years)

40

Plant footprint (m2)

36 000 (CAREM25)

RPV height/diameter (m)11 / 3.2 (CAREM25)

RPV weight (metric ton)267 (CAREM25)

Seismic Design (SSE)

0.25g (CAREM25)

Fuel Cycle Requirements or Approach

390 full-power days and
50% of core replacement(CAREM25)
Core heat removal by
natural circulation, pressure
suppression containment
Under construction(CAREM25)

Distinguishing features

Design status

284 / 326 (CAREM25)
UO2 pellet/hexagonal

1. Introduction
CAREM is a national SMR development project, based on LWR technology, coordinated by Argentina’s
National Atomic Energy Commission (CNEA) in collaboration with leading nuclear companies in Argentina
with the purpose to develop, design and construct innovative small nuclear power plants with high level ofsafety and economic competitiveness. CAREM is an integral PWR type NPP, based on indirect steam cyclewith features that simplify the design and support the objective of achieving a higher level of safety.
CAREM25 is the demonstration prototype of CAREM SMR, and was developed using domestic technology,
at least 70% of the components and related services for CAREM were sourced from Argentinean companies.

2. Target Application
CAREM is designed as an energy source for electricity supply of regions with small demands. It can also
support seawater desalination processes to supply water and energy to coastal sites.

3. Design Philosophy
CAREM is a natural circulation based indirect-cycle reactor with features that simplify the design and improvesafety performance. Its primary circuit is fully contained in the reactor vessel and it does not need any primary
recirculation pumps. The self-pressurization is achieved by balancing steam production and condensation in11

the vessel, without a separate pressurizer vessel. CAREM design reduces the number of sensitive components
and potentially risky interactions with the environment. Some of the significant design characteristics are:Integrated primary cooling system; Self-pressurized; Core cooling by natural circulation; In-vessel control rod
drive mechanisms; and Safety systems relying on passive features.

4. Main Design Features(a) Nuclear Steam Supply System
CAREM is an integral reactor. Its high-energy primary system (core, steam generators, primary coolant andsteam dome) is within a single pressure vessel. Primary cooling flow is achieved by natural circulation, whichis induced by placing the steam generators above the core. Water enters the core from the lower plenum. Afterbeing heated, the coolant exits the core and flows up through the chimney to the upper steam dome. In theupper part, water leaves the chimney through lateral windows to the external region. It then flows down throughmodular steam generators, decreasing its enthalpy.

(b) Reactor Core
The reactor core of CAREM25 has hexagonal cross section fuel assemblies. There are 61 fuel assemblies with
1.4 meters active length. Each fuel assembly contains 108 fuel rods with 9 mm outer diameter, 18 guidethimbles and one instrumentation thimble. The fuel is 1.8% - 3.1% enriched UO2. The fuel cycle can be tailoredto customer requirements, with a reference design for the prototype of 390 full-power days and 50% of corereplacement.

(c) Reactivity Control
Core reactivity is controlled using Gd2O3 as burnable poison in specific fuel rods and movable absorbing
elements belonging to the power adjustment and reactivity control system. Neutron poison in the coolant is
not used for reactivity control during normal operation and in reactor shutdown. Each absorbing element
consists of a cluster of rods linked to a structural element (‘spider’), so the whole cluster moves as a single
unit. Absorber rods fit into the guide tubes.

(d) Reactor Pressure Vessel and Internals
Reactor Pressure Vessel (RPV) of CAREM25 is 11
meters high and 3.4 meters in diameter, with a variable
thickness of 13cm to 20cm. The RPV is made of forgedsteel with an internal stainless steel liner.

(e) Reactor Coolant System
The reactor´s coolant system of CAREM is fully
contained in the RPV. Core cooling is achieved using
natural circulation phenomenon with no primary coolant
pumps. The reactor´s core heats the coolant causing it toflow upwards through the riser. Then it turns downwards
to flow into the steam generators, reducing its
temperature. Finally, it flows to the downcomer and thecore, closing the circuit.

CAREM25 RPV fabrication(f) Steam Generator
In CAREM25, twelve (12) identical mini-helical verticalsteam generators of the once-through type are placed
equidistant from each other, along the inner surface of
the RPV. Each consists of a system of 6 coiled piping
layers, 52 parallel pipes of 26 m active length. They areused to transfer heat from the primary to the secondary
circuit, producing superheated dry steam at 4.7MPa. The
secondary system circulates upwards within the tubes,
while the primary coolant moves in counter-current flow.
The steam generators are designed to withstand theprimary pressure. The entire secondary side is designed
to withstand primary pressure up to isolation valves(including the steam outlet/water inlet headers) in caseof SG tube breakage.

CAREM25 steam generator fabrication(g) Pressurizer
Self-pressurization of the primary system in the steam dome is the result of the liquid–steam equilibrium. The
large steam volume in the RPV, acting as an integral pressurizer, also contributes to damping of any pressure
perturbations. Due to self-pressurization, the bulk temperature at core outlet corresponds to saturation
temperature at primary pressure. In this way, typical heaters present in conventional PWR pressurizers are12eliminated.

5. Safety Features(a) Engineered Safety System Approach and Configuration
Defence in Depth (DiD) principle was the basis of this process. Appropriate criteria were defined for DiD
internalization in the design taking into account CAREM general design characteristics, to prevent, control
and mitigate the postulated events. Western European Nuclear Regulators Association proposal for DiD levels
definition was adopted, which allows the consideration of Multiple Failures Events (Level 3B) as part of DiD
Level 3 for the prevention of core damage. A strategy was defined for each one of the levels. For Level 2 theCVCS can control a medium loss of coolant event and LOHS. Two stages for level 3A and level 3B are
established. For level 3A, the first stage is accomplished by means of passive safety systems, while the second
stage is implemented by active systems in order to achieve the final safe state. For Level 3B, which is used incase of failure in Level 3A, the first stage is also accomplished by systems with passive processes while thesecond stage is composed by simple systems with external water supply. For Level 4 -postulated severe
accident mitigation- provisions are considered to guarantee hydrogen control and RPV lower head coolingwith the aim of in-vessel corium retention.
CAREM´s safety system consists of two reactor protection systems (RPS), two shutdown systems, passive
residual heat removal system (PRHRS), safety and depressurization valves, low pressure injection system anda containment of pressure suppression type. The First Shutdown System (FSS) consists of 9 fast shutdown
rods and 16 reactivity adjust and control rods located over the core, which fall by gravity when needed. While
the Second Shutdown System consists of a gravity assisted high-pressure injection of borated water from twohigh pressured tanks, which is activated automatically when failure of the FSS is detected.
The different systems, structures and components (SSCs) that contribute to the reactor´s safety are classified
according to the identification of low-level safety functions (LLSF) -derived from the fundamental safetyfunctions- and safety functional groups of SSCs that fulfill those functions.

(b) Decay Heat Removal System
During the grace period of 36 hours, core decay heat removal is assured by one out of two PRHRS in the caseof loss of heat sink (LOHS) or Station Black-out (SBO). SBO is considered to be a design basis event. Afterthat period, redundant active systems provide heat removal from the RPV and the containment to the final heatsink. Two redundant diesels provide energy supply for these systems. Despite the low frequency of an SBO
longer than 36 hours, provisions for that scenario are supplied by simple systems supported by a fire
extinguishing system or self-powered pumps. The PRHRS are heat exchangers formed by parallel horizontal
U-tubes (condensers) coupled to common headers.

(c) Emergency Core Cooling System
In case of a loss of coolant event, a Medium Pressure Injection System delivers water into the RPV to keep thecore covered during the grace period. This system consists of two redundant borated water accumulatorsconnected to the RPV. When the pressure in the reactor vessel becomes relatively low, the rupture disks, thatisolate the accumulator tanks from the RPV will break. After the grace period and in case the loss of the coolant
could not be isolated, an active system provides long term water injection into the RPV. Two redundant diesels
provide energy supply for these systems.

(d) Spent Fuel Cooling Safety Approach/System
In case of a SBO and during the grace period, longer than the reactor one, the water of the spent fuel pool is
the heat sink. After this period, a chain of redundant active systems provides heat removal to the final heatsink. Two redundant diesels provide energy supply for these systems.

(e) Containment System
The cylindrical containment vessel with a pressure suppression pool is a 1.2m thick reinforced concrete
external wall with a carbon steel liner and withstands earthquakes of 0.25g. It is designed to withstand thepressure of 0.5MPa. The heat sink is located inside the containment, this provides protection for extreme
external events during the grace period. After grace period, redundant active systems with diesel generator
support, will provide suppression pool cooling and containment depressurization.

6. Plant Safety and Operational Performances
The natural circulation of coolant produces different flow rates in the primary system according to the power
generated and removed. Under different power transients, a self-correcting response in the flow rate is
obtained. Due to the self-pressurizing of the RPV, the system keeps the pressure close to the saturationpressure. The control system is capable of keeping the reactor pressure practically at the operating set point
through different transients, even in case of power ramps.

7. Instrumentation and Control Systems
Plant control is performed by a distributed control system, computer based and with high availability. There
13

are two diverse protection systems: First Reactor Protection System (FRPS) and Second Reactor Protection
System (SRPS), each system carries four redundancies. There are two (2) diverse nuclear instrumentation
systems (NIS), one each for the FRPS and SRPS.

8. Plant Layout Arrangement

CAREM plant layout

CAREM25 construction site

9. Testing Conducted for Design Verification and Validation
A full-scale loop has been built in order to test the innovative control rod drive mechanisms. This facility is
operating at the same parameters (pressure and temperature) as RPV plant design conditions and is was
designed for reactivity adjustment and control rods calibration.

10. Design and Licensing Status
CAREM25 basic design is completed. CAREM25 detail design is 90% progress. CAREM25 was granted aconstruction license. CNEA is working on the commissioning license request.

11. Fuel Cycle Approach
The fuel cycle can be tailored to customer requirements, as a reference CAREM25 has an open fuel cycle
design of 390 full-power days and 50% of core replacement.

12. Waste Management and Disposal Plan
Waste management facilities for waste treatment and storage are provided. Provisions for prolonged temporarystorage of solid wastes at the site are considered.

13. Plant Economics
As CAREM25 is a prototype, plant economics is not provided. CNEA is working on economic studies for thecommercial module.

14. Development Milestones
1984
2006
2009
2011
2011
2012
2013
2014
2024
2026

14

CAREM concept was presented in Lima, Peru, during the IAEA Conference on SMRs
and was one of the first of the new generation reactor designs. CNEA officially launched
the CAREM project
Argentina Nuclear Reactivation Plan listed the CAREM25 project among priorities ofnational nuclear development
CNEA submitted its preliminary safety analysis report (PSAR) for CAREM25 to the
ARN
Start-up of a high pressure and high temperature loop for testing the innovative hydrauliccontrol rod drive mechanism
Site excavation work beginning
Civil design beginning
Construction license
Civil works started
Electromechanical erection (start)
First criticality

ACP100 (CNNC/NPIC, China)
All content included in this section has been provided by, and is reproduced with permission of, CNNC/NPIC, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of
CNNC/NPIC, China
originReactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Integral PWR
Light water / light water
385 / 125

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15 / 4.6

Core inlet/outlet coolant
temperature (oC)

286.5 / 319.5

Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup (GWd/ton)

UO2 / 17 × 17 square pitch
57

Reactivity control mechanism

Control rod drive mechanism(CRDM), Gd2O3 solid burnablepoison and soluble boron acid
Passive60
200 000
10 / 3.35
300
0.3 g
Temporarily stored in spent fuelpools
Integrated reactor with tube-in-tube
once through steam generator,
nuclear island undergroundConstruction in progress

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements/approachDistinguishing features

Design status< 4.9524< 52

1. Introduction
The ACP100 is an integrated PWR design developed by China National Nuclear Corporation (CNNC) togenerate an electric power of 125 MW(e). The ACP100 is based on existing PWR technology adapting
verified passive safety systems to cope with the consequences of accident events; in case of transients andpostulated design basis accidents the natural convection cools down the reactor. The ACP100 integrated
design of its reactor coolant system (RCS) enables the installation of the major primary circuit’s components
within the reactor pressure vessel (RPV).

2. Target Application
The ACP100 is a multipurpose power reactor designed for electricity production, heating, steam production
or seawater desalination and is suitable for remote areas that have limited energy options or industrial
infrastructure.

3. Design Philosophy
The ACP100 realizes design simplification by integrating the primary cooling system and enhanced safety
by means of passive safety systems.

15

4. Main Design Features(a) Nuclear Steam Supply System
The integrated nuclear steam supply system (NSSS) design consists of the reactor core, and sixteen (16)once-through steam generators (OTSG). The four (4) canned motor pumps are installed nozzle to nozzle to
the RPV.

(b) Reactor Core
The 57 fuel assemblies (FAs) of ACP100 core with total length of 2.15 m core have a squared 17 × 17
configuration. The fuel 235U enrichment is about 1.9 – 4.95%. The reactor will be able to operate 24 months
at balance fuel cycle.

(c) Reactivity Control
The reactivity is controlled by means of control rods, solid burnable poison and soluble boron dissolved in
the primary coolant. There are 20 control rods, with a magnetic force type control rod driving mechanism(CRDM).

(d) Reactor Pressure Vessel and Internals
The RPV and equipment layout are designed to enable the natural circulation between reactor core and steam
generators. The RPV is protected by safety relief valves against over-pressurization in the case of strong
difference between core power and the heat removed from the RPV. The internals not only support andfasten the core but also form the flow path of coolant inside RPV.

(e) Reactor Coolant System
The ACP100 primary cooling mechanism under normal operating condition and shutdown condition is doneby forced circulation. The RCS has been designed to ensure adequate cooling of reactor core under all
operational states, during and following all postulated off normal conditions. The integral design of RCS
significantly reduces the flow area of postulated small break LOCA.

(f) Steam Generator
There are 16 OTSGs, which are mounted within the RPV. All the 16 OTSGs are fitted in the annulus
between the reactor vessel and hold-down barrel. The bottoms of OTSGs are limited their position by thehole on barrel supporting hub, the heads are welded to the reactor vessel steam cavity.

(g) Pressurizer
The pressurizer of ACP100 is located outside of the reactor vessel. The pressurizer is a vertical, cylindrical
vessel with hemispherical top and bottom heads.

(h) Primary pumps
The ACP100 uses canned-motor pumps as reactor coolant pumps which are directly mounted on the RPV
nozzle. The shaft of the impeller and rotor of the canned-motor pump is contained in the pressure boundary,eliminating the seal LOCA of reactor coolant pump.

5. Safety Features(a) Engineered Safety System Approach and Configuration
The ACP100 is designed with several passive safety features and severe accident mitigation features.
Enhanced safety and physical security of ACP100 are made possible by arranging the nuclear steam supply
system and spent fuel pool underground. When the spent fuel pool is filled with spent fuel of 10 years, thepool cooling water can cope for seven (7) days of cooling in the case of accident before boiling dry anduncovering fuel. Severe accident prevention and mitigation are achieved through passive reactor cavity
flooding preventing RPV melt, passive hydrogen recombination system preventing containment hydrogen
explosion and maintaining the containment integrity after severe accidents, automatic pressure relief system
and RPV off-gas system to remove non-condensable gas gathered at RPV head after accidents.

(b) Decay Heat Removal System
The PDHRS prevents core meltdown in the case of design basis accident (DBA) and beyond DBA, such as
station black out, complete loss of feedwater, small-break LOCA (i.e., to prevent the change of beyond DBA
to severe phase). The PDHRS of ACP100 consists of one emergency cooler and associated valves, piping,and instrumentation. The emergency cooler is located in the in-containment refuelling water storage tank,which provides the heat sink for the emergency cooler. The decay heat is removed from the core by natural
16

circulation. The PDHRS provides core cooling for seven (7) days without operator intervention or long termwith IRWST makeup water collected by gravity force from the steam condensed in containment.

(c) Emergency Core Cooling System
The emergency core cooling system (ECCS) consists of two coolant storage tanks (CST), two safety
injection tanks (SIT), an in-containment refuelling water storage tank (IRWST) and associated injection lines.
The ACP100 has a safety related direct current (DC) power source to support accident mitigation for up to
72 hours, along with auxiliary power units to recharge the battery system for up to seven (7) days. After
LOCA accidents, the steam in containment is condensed continuously at containment internal face thus theheat is conducted to containment, which is cooled by PAS, thus ensuring the containment integrity.

(d) Spent Fuel Cooling Safety Approach / System
The In-Refuelling Water Storage Tank (IRWST) is a passive water tank, resting on the internal structure baseslab. During refuelling operations, it provides water for refuelling cavity, internals storage compartment andrefuelling transfer canal to complete the refuelling operation. Under the condition of LOCA and the steam
pipe rupture, it provides water for emergency reactor core cooling. In the severe accidents, water in it floods
the internal structure under the balanced water level due to gravity. During the operation of reactor automatic
depressurization system, it absorbs the sprayed steam from the RCS. During the operation of the passive
residual heat removal cooler, it works as the heat sink of the passive residual heat removal system. The
reactor pool is used during refuelling operation or inspection of reactor internals. The reactor pool consists of
two compartments which can be separated by bulkhead: reactor cavity and internals storage compartment
adjacent to the reactor.

(e) Containment System
The ACP100 containment houses the RCS, the passive safety systems and the auxiliary systems. ACP100
adopts small steel containment cooled by air with no need of drive signal.

6. Plant Safety and Operational Performances
Nuclear safety is always the first priority. The ultimate goal of nuclear safety is to establish and maintain an
effective defence that can effectively protect people, community, and environment from radioactive disaster.
To be specific, the design and operation of ACP100 ensures that radiation dose to the workers and to themembers of the public do not exceed the dose limits and kept it as low as reasonably achievable. Accident
prevention measures ensure that radioactive consequences are lower than limited dose in terms of all theconsidered accident sequences and even in the unlikely severe accidents, mitigation of accidents induced
influences can be ensured by implementing emergency plan. The design of ACP100 incorporates operational
experience of the state-of-the-art design. Proven technology and equipment are adopted as much asreasonably possible.

7. Instrumentation and Control System
The Instrumentation and Control (I&C) system designed for ACP100 is based on defence in depth concept,
compliance with the single failure criterion and diversity. The diversity in the design of I&C system isachieved through: (1) different hardware and software platforms for 1E and N1E I&C, (2) reactor protection
system (RPS) with functional diversity, and (3) diverse protection systems to cope with the common mode
failure of the RPS. I&C systems of the NSSS include reactor nuclear instrumentation system, RPS, diverse
actuation system, reactor control system, rod control and rod position monitoring system, reactor in–core
instrumentation system, loose parts and vibration monitoring system and other process control systems.

8. Plant Layout Arrangement
The ACP100 adopts compact single-unit plant layout，
including nuclear island building (NI) and turbine
generator building (CI). The fuel building, electricalbuilding and the nuclear auxiliary building arearranged around the reactor building, allowing the NI
building to work well with a smaller size. This layout
can adjust itself well to various kinds of plant sites.
The operation platform of the reactor, the operation
platform of fuel and the transportation platform of theradioactive waste are arranged around the ground of
the power plant, which simplify the transportation of
the fuel, radioactive waste and big equipment in theNI, to lower the using frequency of NI hoists as wellas the cost of construction.
Plant layout arrangement

17

The turbine generator building (CI) is arranged longitudinal to the main nuclear building. The head of steam
turbine faces towards the nuclear building. The moisture separator re-heater (MSR) is arranged on the otherside of operation layer of high-pressure cylinder. The plant is mainly equipped with turbine, generator,
excitation device, MSR, condenser, condensate pump, low-pressure heater, deaerator, feed pump and otherauxiliary equipment.

9. Testing Conducted for Design Verification and Validation
Seven test research of verification tests have been finished, such as Control rod drive line cold and hot test,
Control rod drive line anti-earthquake test, Internals vibration test research, Fuel assembly critical heat flux
test research, Passive emergency core cooling system integration test, CMT and passive residual heatremoval system test research, Passive containment heat removal testing.

10. Design and Licensing Status
The ACP100 preliminary safety assessment report (PSAR) is approved by National Nuclear Safety Authority
and detailed engineering design is in progress. Changjiang nuclear power site, Hainan, China, was chosen tobuild the first of a kind (FOAK) ACP100 demonstration project. FCD in July 2021.Construction period of
FOAK 55 months, target commercial operation in 2026.

FCD in July 2021

11. Fuel Cycle Approach
Spent fuel processing is similar to other nuclear power plants. It is temporarily stored in spent fuel pools.
Waste management approach and disposal plan is similar to other nuclear power plants.

12. Waste Management and Disposal Plan
Waste management approach and disposal plan is similar to other nuclear power plants.

13. Development Milestones
2016
2017

2018
2019
2020
2021
2026

18

Generic reactor safety review for ACP100 by IAEA finished.
CNNC signed an agreement with the Changjiang municipal government in Hainan
Province to host the first of a kind (FOAK) ACP100 demonstration unit.
Preliminary safety assessment report (PSAR) finished.
PSAR submitted to National Nuclear Safety Authority, Site Preparation started.
Apply for authorize to Changjiang nuclear power site, Hainan, China.
FCD on Changjiang nuclear power site, Hainan, China.
Target commercial operation.

CANDU SMRTM (Candu Energy Inc,
Canada)All content included in this section has been provided by, and is reproduced with permission of Candu Energy Inc, Canada

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

Candu Energy Inc. Member of the
SNC-Lavalin Group, Canada
Pressurized Heavy Water Reactor

Coolant/moderator

Heavy Water (D2O)

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

960 / 300

NSSS Operating Pressure(primary/secondary), MPa
Core Outlet Coolant
Temperature (oC)
Fuel type/assembly array

9.9 / 4.6

Number of fuel assemblies in thecore
Fuel enrichment (%)

2 064

Core Discharge Burnup (GWd/ton)

5.8

Refuelling Cycle (months)

On-lineReactivity control mechanism

Zone controllers, mechanical adjusters

Approach to safety systems

Combined active and passive

Design life (years)

70 years

Plant footprint (m2)

21 000RPV height/diameter (m)NA; Calandria

RPV weight (metric ton)

300

Seismic Design (SSE)

0.3g

Fuel cycle requirements / Approach

Natural Uranium, once-throughDistinguishing features

Natural Uranium fuel (noenrichment); high degree of
localization
Conceptual design

Design status

Forced310
37 elements

Natural Uranium; not enriched

1. Introduction
The CANDU® Small Modular Reactor (SMR) is being developed as a proven and mature design to help
countries reach their goal of Net Zero. Built on proven CANDU® technology to be quickly deployable, this
300 MW(e) reactor features simplified systems, fewer components and a modular design. The design
objectives are a low-cost, low-carbon power with a high capacity factor in a compact layout.

2. Target Application
A CANDU SMR is designed to enable a fast deployment using proven technology, maintaining energy
independence by using natural uranium fuel from fuel manufacturers and avoiding the need to import enricheduranium fuel. This maximizes identification and utilization of a high performing supply chain, minimizing
project delivery risk and creating high technology jobs in countries. The CANDU SMR (CSMR) offers asmaller and more flexible source of carbon-free electricity. The CSMR is a Generation III+ reactor with adesign life of 70 years (extendable to 100 years) and a 90% capacity factor. It accommodates to a broad rangeof potential reactor sites, with its 0.3g seismic design, compact shape and grid-stability features.

3. Design Philosophy
The CSMR is based on a decades-long proven design that is licensed in many countries operating CANDU
reactors. It meets modern regulatory requirements, with dedicated post-Fukushima features.
The CSMR does not require enriched fuel or a fuel-qualification program because it relies on natural uranium
in the same fuel bundle employed in multiple CANDU6® reactors. To achieve this it uses heavy water for19moderation and cooling.

4. Main Design Features
Nuclear Steam Supply System
The CSMR is a horizontal pressure tube, pressurized heavy water reactor developed from the long lineage ofsuccessful CANDU reactors. The reactor core consists of a horizontal calandria housing a set of pressure
tubes. The reactor coolant system consists of these pressure tubes, two steam generators, four primarycirculation pumps plus interconnecting piping and headers. The calandria itself is filled with heavy water thatsurrounds the pressure tubes and provides neutron moderation plus an element of safety, with the calandria
operating as a core catcher for severe accidents.

Reactor Core
The CSMR core is 176 channels in a lattice of 14 rows and 14 columns. The channel array is 12’ 2” across.
The calandria main shell radius is 285 cm. Reactor is fueled by standard CANDU natural uranium fuel bundle.

Reactivity Control
Reactor control uses eight zone control units to provide the primary means of reactivity regulation during
normal operation. The core design allows control to be achieved through very small changes in these zone
control units. The adjuster rods, absorber rods and ability to use soluble poisons found in traditional CANDU
designs are retained, as are two fully independent, diverse safety-shutdown systems: SDS1 and SDS2. Like
its predecessors, the CSMR provides safety shutdown reliability.

Reactor Pressure Vessel and Internals
The reactor consists of a cylindrical calandria and end shield assembly that is enclosed and supported by thecylindrical shield tank and its end walls. The cylindrical shield tank extension assembly closes the top of both
vessels. The calandria contains the heavy water moderator and reflector; the shield tank contains light water.
The seismically-qualified moderator system is independent from the pressurized heavy water heat transportsystem (HTS) in the fuel channel assemblies and has the inlet and outlet nozzles connected high in the calandria
to enhance performance in the event of a severe accident.

Reactor Coolant System
The HTS is a single figure-of-eight loop with two steam generators and four heat transport pumps. The HTS
is seismically qualified such that a seismic-induced LOCA is outside of the design basis.

Steam Generator
Two steam generators are used, both with inverted U-tubes and integral steam drum and preheaters, and locatedinside the containment structure.

Pressurizer
The pressure and inventory control system include a pressurizer, bleed condenser, feed pumps (one operating;one on standby), a storage tank and control valves.

5. Safety Features
The CSMR safety features establish defence-in-depth against radiological hazards. CSMR leverages theinherent safety characteristics of the basic CANDU reactor design and supplements them with a judicious
application of passive and active safety features that emphasizes provenness and results in an improvement insafety. The irradiated fuel bay is a robust, seismically qualified structure with a large volume of water relative
to the decay heat load of discharged fuel, providing many days of passive cooling in the event of a loss ofactive heat removal.
The CSMR provides large volumes of water that are available to provide cooling to the core in the event ofaccidents, including by passive means. In addition, it has a large containment volume, contributing tominimizing hydrogen concentrations in severe accidents. The CSMR HTS is seismically qualified to the designbasis earthquake. The CSMR also makes use of loop subdivision and feeder interlacing to reduce the rate ofcoolant voiding that is possible in the event of a large HTS pipe break.
The CSMR places all reactivity devices in the separate, low-temperature, low-pressure moderator, eliminatingpressure-driven ejection of reactivity devices from the design. The separation of moderator from coolant also
provides two separate heat removal means in the event of accidents and ensures that moderator temperature
feedback to the core physics is negligible in normal operation. The characteristic CANDU pressure-tube designmeans that direct containment heating type of severe accident does not occur in this design.

Engineered Safety System Approach and Configuration
All plant systems are assigned to one of two groups (group 1 or group 2) according to the CSMR grouping andseparation approach. Group 1 includes the power production systems and delivers safety functions for group
2 support. Group 2 includes the safety related systems required for mitigation of accidents and group 2 systems

20

are qualified or protected to provide safety functions in the event of severe external events. There is additional
separation within the groups.
The CSMR has two separate shutdown systems. These are two fully-capable fast-acting means of shutdown
for use at the third level of defence in depth, fully independent of each other and of the reactor regulating
system which acts at the second level of defence in depth.

Decay Heat Removal System
Decay heat removal is accomplished in CSMR by application of several provisions in the design, including
three group 2 systems, as follows:
The large inventory in the HTS, including the pressurizer inventory, provides heat removal for normal
operating transients. The HTS layout enhances natural circulation to the steam generators, which have a large
inventory. The steam generators have normal make-up capability and back-up feedwater.
When the steam generators are not available, or are not effective for heat removal, heat can be removed fromthe HTS using the shutdown cooling system. Under emergency conditions this system can be valved in at full
HTS pressure and temperature.

Emergency Core Cooling System
The emergency core cooling system supplies emergency coolant to the reactor headers in the event of a lossof-coolant accident (LOCA). The system operation is divided into two parts, short-term injection and longterm recirculation. Short-term injection consists of two stages: high pressure and low pressure injection. During
the high pressure injection stage, water from the accumulator tanks is injected into the HTS by pressurizedgas. After this water is depleted, low pressure injection automatically takes over, injecting water from a gradelevel tank via the emergency core cooling pumps. A connection is provided to this tank for demineralized
water makeup, and for initial filling. For small LOCAs provision of steam generator crash cooldown and themaintenance of continued feedwater flow is used instead of the ECC heat exchangers to provide cooling. The
crash cooldown is performed via the MSSVs.

Containment System
The basic function of the containment system is to form a continuous, pressure-confining envelope about thereactor core and primary cooling system in order to limit the release of radioactive material to the external
environment resulting from an accident. This accident could be either a failure of fuel cooling, or an accident
which releases radioactive material into the containment without a rise in containment internal pressure.
To achieve this overall function, the containment system includes the following related safety functions:i. Isolation: to ensure closure of all openings in containment when an accident occurs.
ii. Pressure/activity reduction: to control and assist in reducing the internal pressure and free radioactive
material released into containment by an accident.
iii. Hydrogen control: to limit concentrations of hydrogen/deuterium within containment after an accident toprevent detonation.
iv. Monitoring: to monitor conditions within containment and the status of containment equipment, before,
during and after an accident.
In addition to its safety role, the containment structure also serves the following functions:i. To limit the release of radioactive materials from the reactor to the environment during normal operations.
ii. To provide external shielding against radiation sources within containment during normal operations andafter an accident.
iii. To protect reactor systems against external events such as tornados, floods, etc.
The containment system includes a reinforced concrete containment structure (the reactor building) with areinforced concrete dome and an internal steel liner, access airlocks, equipment hatch, building air coolers forpressure reduction, and a containment isolation system.

6. Plant Safety and Operational Performances
The CSMR design aligns with the concept of defence in depth, leveraging inherent safety features through theapplication of proven engineered systems with an emphasis on providing high reliability through a prudent
mix of active and passive features. The overall CSMR design philosophy is to reduce total unit energy cost by
reducing specific capital cost, shortening the construction schedule, reducing operating, maintenance andadministration costs and providing for plant life extension. In addition, CSMR enhances or improves the
traditional CANDU advantages including real safety, low man-rem exposure, high capacity factor and ease ofmaintenance. Proven systems, system parameters, components, and concepts are used, including proventechnologies from other industries.

7. Instrumentation and Control Systems
The overall I&C architecture and design is aligned with the five levels of defense-in-depth, with suitable
independence between the different levels. Most automated plant control functions are implemented in amodern distributed control system (DCS) using a network of modular, programmable controllers that
communicate with one another using reliable, high-security data transmission methods. The plant is automatedto the extent that requires a minimum of operator actions for all phases of operation.
21

The control systems are backed up by the safety systems, which include the two independent shutdown
systems, the emergency core cooling system, emergency heat removal system, and the containment system.
Each of these safety systems operates completely independent of the other and independent of the reactor andprocess control systems.

8. Plant Layout Arrangement
The CSMR layout is arranged to minimize and achieve a short practicalconstruction schedule. This is achieved by simplifying the layout,
optimizing interfaces, reducing construction congestion, providing accessto all areas, providing flexible equipment installation sequences andreducing material handling requirements. The principal structures of the
CSMR include the reactor building, the reactor auxiliary building, theturbine building, the group 1 service building and maintenance building.
The auxiliary structures include the Group 1 pump house (main pumphouse), Group 2 pump house and administration building. In general,
Group 1 systems sustain normal plant operation and power production,
while Group 2 systems have a safety or safety support functions.

9. Design and Licensing Status

Plant Layout

Candu Energy Inc. has requested the CNSC to initiate the formal Vendor Design Review (VDR) process for
the CSMR design, which will be conducted based on the regulatory requirements applicable to the new build
projects in Canada. The design of CSMR is at the Conceptual Design stage.

10. Fuel Cycle Approach
The standard CSMR uses a once-through fuel cycle based on natural uranium, producing a very low residualuranium spent fuel with low heat generation. It is also capable of burning recovered uranium from light-water
reactors, making it a valuable addition to any light-water fleet. The use of MOX and Thorium fuels are also
possible with suitable customization.

11. Waste Management and Disposal Plan
The CSMR implements systems and equipment for handling radioactive wastes consistent with the state-ofthe-art in Canada. The radioactive waste management systems are designed with the objective of limiting
routine releases in accord with the ALARA principle. Facilities are provided for interim storage, or controlled
release, of all radioactive gaseous, liquid and solid wastes.
It is anticipated that most of the of low-level and intermediate-level solid waste produced over the reactor’s
70-plus year lifetime will be stored at the waste management area, with high-level waste being stored on-sitein a dry storage facility. Candu Energy Inc. is working on establishing the plan to safely store, handle anddispose of all the spent fuel including the on-site storage, long term storage and potential design for new longterm storage containers that meet the requirements for the deep geological repository design of the Nuclear
Waste Management Organization (NWMO).

12. Development Milestones
2020
CED + 1
CED + 3
CED + 5
CED + 822

Design Concept ready for contracting
Conceptual Design complete (CED = Contract Effective Date)
Detailed Design complete
First Concrete for first unit
First Unit in service

CAP200 (SPIC/SNERDI, China)
All content included in this section has been provided by, and is reproduced with permission of, SPIC/ SNERDI, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
SPIC/SNERDI, China
of origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)PWR
Light water / Light water
660 / >200

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15.5 / 6.06

Core inlet/outlet coolant
temperature (oC)289 / 313

Fuel type/assembly array
Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)
Reactivity control mechanism

UO2 pellet / 17x17 square

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approachDistinguishing features
Design status

89
4.2 (Average)
24
37 (Average)
Control rod drive mechanism andsoluble boron
Passive60–
8.845 / 3.280
200
0.30 g
UO2 fuel and low-leakage core design
Compact layout; Passive
Underground containment
Basic design

safety;1. Introduction
The China Advanced Passive pressurized water reactor 200MW(e) (CAP200) is one of the serial researchand development products of PWRs adopting passive engineered safety features initiated by Shanghai
Nuclear Engineering Research and Design Institute (SNERDI). The design of CAP200 is based on more than
45 years of experience in PWR R&D, and more than 20 years of experience in PWR construction and safe
operation in China. It is the outcome of accumulated experience and achievements of the world's first batchof AP1000 units and the R&D of CAP1400. Furthermore, it adopts safety enhancement measures based on
lessons learnt from the Fukushima event.

2. Target Application
CAP200 is designed for multiple applications, such as nuclear cogeneration and replacing retired fossil
power plants in urban areas. It can be used as a supplement to large PWRs.

3. Design Philosophy
CAP200 is a small PWR which is designed with improved safety, flexibility and environmental friendliness,and is also comparable with other SMRs on economy. Compared with large PWRs, CAP200 has a number ofadvantages such as higher inherent safety, lower frequency of large radioactivity release, longer time of postaccident no operator intervention, smaller environmental impact, lower site restrictions, shorter construction
period and smaller financing scale as well as lower financial risk. The main features of this reactor are as
follows:

23

- Compact primary system: Steam generators (SGs) are connected to reactor pressure vessel (RPV) directly
and main pipes are eliminated. Compact layout of system and components results in lower risk andpossibility of loss of coolant accidents and smaller primary system footprint.
- Modular design and fabrication: main modules can be fabricated in factory and transported to the site for
installation. Construction period can be shortened because of high modularization.
Redundant and diverse safety features: redundant and diverse active and passive safety features are deployed,
which ensures the reactor core safety and extremely low risk of large radioactivity release.

4. Main Design Features(a) Nuclear Steam Supply System
The NSSS of CAP200 consists of reactor pressure vessel (include IHP), steam generators, reactor coolant
pumps, pressurizer and auxiliary systems. Due to the cancellation of RCS pipe, the radioactivity containment
capability of CAP200 is better than traditional PWR.

(b) Reactor Core
The CAP200 reactor core adopts two different types of control rod assemblies. There are 37 assemblies intotal. The high worth assemblies are known as “black” rods. These are used for shutdown and large swingsof reactivity. The grey rods with lower worth are used for load-follow during operation of a fuel cycle to
avoid adjusting soluble boron concentration, which results in a substantial reduction in wastewater
generation and treatment for PWR required to execute load-follow operation.

(c) Reactivity Control
Core reactivity is controlled by both soluble boron and control rods. CAP200 is capable of load following
without boron dilution. Nevertheless, the control rods need repositioning when boron dilutes. This method
suppresses the excess reactivity with soluble boron as conventional PWR does, but to a large degree
simplifies the conventional boron system and the dilution operation.

(d) Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) of CAP200 is designed with reference to that of CAP1400. The difference
is in the inlet and outlet of RPV. The pressure nozzles are used in the case of CAP200, to connect RPV and
SGs replacing main pipes. The vessel is cylindrical with a removable flanged hemispherical upper head andhemispherical bottom head. The area with highest neutron fluence of active core region of the vessel iscompletely forged and free of welding which enhances the confidence of 60 years’ design life and decreases
the task of in-service inspection. As a safety enhancement, neutron and temperature detectors enter thereactor through upper head of reactor vessel which eliminates penetrations in the lower head of RPV. This
reduces the possibility of a LOCA and uncovering of the core The core is positioned as low as possible in thevessel to decrease re-flood time in the event of an accident. Furthermore, this arrangement is very helpful for
successful execution of IVR. The reactor internals, installed in RPV, provide support, protection, alignment
and position for the core and control rods to guarantee safe and reliable operation of reactor. The reactor
internals consist of the lower internals and the upper internals. The core shroud made of stainless steel iswelded which eliminates the occurrence of reactor core damage induced by loosening of baffle bolts. The
reactor internals design of CAP200 takes the inner pressure nozzle installation into consideration.

(e) Reactor Coolant System
The primary circuit is designed to achieve a high safety and performance record. The system consists of twoheat transfer circuits, each with a steam generator and a reactor coolant pump installed directly onto thesteam generator; eliminating the primary piping between the pump and steam generator, and also the primarypiping between the reactor vessel and steam generator. The reactor vessel and steam generators areconnected directly through the interface nozzles. In addition, the system includes the pressurizer,
interconnecting piping, valves, and instrumentation for operational control and safeguards actuation. A
simplified support structure for the primary systems reduces in-service inspections and improves
accessibility for maintenance. The reactor coolant system pressure boundary provides a barrier against
radioactivity release and is designed to provide a high degree of integrity throughout operation of the plant.

(f) Steam Generator
The steam generator (SG) of CAP200 is vertical U-tube type with 3861 tubes made of thermally treatednickelchromiumiron Alloy 690 on a triangular pitch, to fulfil the requirement of heat transfer capacity. The
design parameter of steam exit from SG is 6.06 MPa and 183.6 kg/s for thermal design flow rate with notube plugged, respectively. Pressure nozzles are used to connect RPV and RCP directly with SG. The innerduct of the pressure nozzle connecting RPV with SG side is welded to the partition board of SG waterchamber. The nozzles are designed with two ducts. For the nozzles between RPV and SGs, hot water flows
in the inner circular duct and cold water in the outer annular duct. Stainless steel bars have been installed in24bend area to preclude occurrence of damaging flow induced vibration under all conditions of operation. The
feedwater spray nozzle is at top of the feed ring and adopts the separated start-up feedwater pipe which
eliminates thermal stratification and prevents occurrence of water hammer. The SG channel head is divided
into three parts, hot channel, cold channel, and a third channel through which the coolant is pumped backinto reactor vessel.

(g) Pressurizer
The pressurizer of CAP200 is a typical steam-type with electrical heater at the bottom and spray at the top.
The pressure control is steady and reliable. The pressurizer is a vertically mounted cylindrical pressure vessel
with hemispherical top and bottom heads which adopts the traditional design that is based on proventechnology. The pressurizer volume is designed to be large which increases margins for transient operationand minimizes unplanned reactor trips and provides higher reliability. Also, fast-acting power-operated relief
valve, which is one of the reactor coolant system leakage sources and the component requiring potential
maintenance, won’t be needed because of the improved transient response by using large volume pressurizer.

(h) Primary pumps
Both leak-tight canned motor pump and wet coil pump are possible choices for the RCP of CAP200. There isrich experience of use of canned motor pump and wet coil pump in nuclear power plants, and the current
advanced large nuclear reactors are also employing canned motor pump or wet coil pump. The RCP is
designed to produce a head of 56.3m at design flow rate of 12000 m3/h with a cold leg temperature of 289℃.
The reactor coolant pump has no shaft seals, eliminating the potential seal failure LOCA, which significantly
enhances safety and reduces pump maintenance. The pumps have an internal flywheel to increase the pumprotating inertia and thereby providing a slower rate-of-flow coastdown to improve core thermal margins
following the loss of electric power.

5. Safety Features(a) Engineered Safety System Approach and Configuration
CAP200 adopts passive safety systems which take advantage of natural forces such as natural circulation,
gravity and compressed air to make the systems work, offering improvements for plant in simplicity, safety,
O&M, availability and investment protection. No active components such as pumps, fans and othermachinery are used. A few simple valves align and automatically actuate the passive safety systems. The
passive safety systems are designed to meet criteria of single failure, independence, diversity, multiplicity.

(b) Passive core cooling system
The main function of passive core cooling system is toprovide emergency core cooling during postulated
design basis accidents by supplement and boration to
RCS after non-LOCA accidents and safety injection tocore after LOCA. Passive core cooling system forms
core decay heat removal pathway together with passive
containment cooling system. For CAP200, Passive
Safety Injection System combined with the Passive
Decay Heat Removal System is referred as the Passive
Core Cooling System.

(c) Spent Fuel Cooling Safety Approach / System
The Spent Fuel Cooling System is not required to operate to mitigate design basis events. In the event the
spent fuel pool cooling system is unavailable, spent fuel cooling is provided by the heat capacity of the waterin the pool. With the use of makeup water from safety related sources, the pool level is maintained above thespent fuel assemblies for at least 72 hours.

(d) Passive Containment Cooling System
The containment of CAP200 is submerged in a water pool. After a steam line break accident or a loss of
coolant accident, heat will be transferred from steam in containment to the water pool. The water pool issafety-related and prevents the containment from exceeding the design pressure and temperature following a
postulated design basis accident by cooling the outside surface of containment, as shown in the above figure.
The inventory in the water pool can last at least 7 days after an accident. The Passive containment coolingsystem works without operator control or external assistance.

6. Plant Safety and Operational Performances
Moderating and maximizing the time response of event loads relative to their limits is a focal point in
improving the reactor inventory and cooling safety functions. The total inventory and its distribution
25

throughout the system factor into this assessment. Further, reserve primary coolant from interfacing safetysystems, most notably the RWST, can extend these time response periods both temporally and to a broader
range of off-normal plant states. The arrangement of reactor core and steam generator thermal centers is
crucial to the plant's capability to remove heat by natural circulation following a loss of forced circulation.
For CAP200, these two components are vertically separating within an integral pressure vessel.

7. Instrumentation and Control System
The instrumentation and control (I&C) system provides the capability to monitor, control and operate plantsystems. It functions to (1) control the normal operation of the facility, (2) ensure critical systems operate
within their designed and licensed limits, and (3) provide information and alarms in the control room for theoperators. Important operating parameters are monitored and recorded, during both normal operations andemergency conditions to enable necessary operator actions. The I&C system is implemented using modern,scalable digital technology.

8. Plant Layout Arrangement
The reactor building is placedunderground, which issuitable for various geological conditions. It canadapt to the site conditions of most potential sites andislands. The size of Nuclear Island is minimized byboth system simplification and adoption of passive
design. Compact RCS especially eliminates main
pipes and allows reactor building to be smaller. The
number of system component is reduced by adoptionof large capacity equipment, common use of single
equipment for different systems, for example,
replacing the polar crane by a travelling crane to use
both for reactor building and auxiliary building, etc.

9. Testing Conducted for Design Verification and Validation
The design of passive safety systems of CAP200 takes advantage of the development results of CAP1400
which has been fully validated by Advanced Core-cooling Mechanism Experiment (ACME) and other tests.
The Integral Effect Test is considered to be developed in order to give further demonstration to theperformance of passive safety systems of CAP200. Hydraulic simulation test of reactor structure and flowinduced vibration test of reactor vessel internals are being implemented to study the uniformity of the inlet
flow distribution of the reactor core, the reactor pressure drops, the bypass flow characteristics and the flowinduced vibration response characteristics due to the hydrodynamic loads and interaction with coolant flow.

10. Design and Licensing Status
Basic design started.

11. Fuel Cycle Approach
There are 89 UO2 fuel assemblies in in the core. This core design assumes an end-of-cycle (EOC) exposureof 703 effective full-power days (EFPD). The feed fuel in this design uses two distinct assembly
enrichments. The selection of the enrichments and placement within the core are meant to minimize neutron
leakage from the core while maintaining acceptable axial and radial power distributions and peaking factors.

12. Waste Management and Disposal Plan
Waste Management System consists of liquid, gaseous, and solid radwaste systems. The liquid radwaste
system is designed to control, collect, process, handle, store, and dispose of liquid radioactive waste
generated as the result of normal operation, including anticipated operational occurrences. The gaseous
radwaste system is designed to receive hydrogen bearing and radioactive gases generated during process
operation and provides the capability to reduce the amounts of radioactive nuclides. The solid radwaste
system is designed to collect, process, package and temporarily store the wastes like spent ion exchange
resins, spent filter cartridges, dry active wastes and etc. After stored for several years, the waste will betransported to the disposal site qualified to receive radioactive waste for geological disposal.

13. Development Milestones
2014
2015
2018
2022

26

Conceptual design startedConceptual design finishedFurther deepen conceptual design
Basic design started

DHR400 (CNNC, China)
All content included in this section has been provided by, and is reproduced with permission of CNNC, China.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

China National Nuclear Corporation(CNNC), China
Pool type reactor

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

400 / no electricity

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

0.3 (Core inlet pressure)

Number of fuel assemblies in
the core
Fuel enrichment (%)

69

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

31

Reactivity control
mechanismApproach to safety systems

Control rod drive mechanisms

Forced circulation

68 / 98
UO2 pellet / 17x17 square＜5

15

Design life (years)

Inherent safety features with large
water inventory in the reactor pool /Passive core residual heat removal
system / Passive heat pipe system
60

Plant footprint (m2)

100 000

Pool depth/diameter (m)28 / 10

Seismic Design (SSE)

0.3g

Distinguishing features

Inherent safety features / Economical/ Heat load tracking capability
Basic design

Design status

1. Introduction
The DHR400 is a pool type reactor dedicated designed for district heating with a thermal power of 400 MW.
The DHR400 operates under low temperature in atmospheric pressure. Established light water reactor
technologies are used in the design of DHR400. Inherent safety features are incorporated to enhance its
safety and reliability. Significant design features include: Large water inventory in the pool that provides
large thermal inertia and a long response time, thus enhances the resistance to system transients andpostulated accidents. The low probability of core meltdown of DHR eliminates the possibility of large-scale
radioactivity release and simplifies the off-site emergency response. Simplification in design andconvenience in maintenance lead to improvement in economic efficiency. With high reliability and inherent
safety features, DHR400 can be located in the vicinity of the targeted heating supply area.

2. Target Application
DHR400 is a pool type reactor for district heating with a thermal power of 400 MW. The reactor is designed
to replace traditional small regional heating plants using coal or natural gas to minimize air pollution andreduce carbon emission.

3. Design Philosophy
The DHR400 is designed based on the pool type research reactor. It operates under low temperature withatmospheric pressure above the pool surface. With the deep reactor pool, appropriate core outlet temperature27

is achieved by the static pressure of the water layer. The design configuration precludes typical reactor
accidents including the loss of coolant accident (LOCA), control rod ejection accident and the loss of decay
heat removal capability. DHR400 is moderated by low pressure water, inherent reactor shutdown at abnormal
situations is assured with the large negative temperature reactivity coefficient. The DHR400 design also
adopts proven pool type reactor technology, design simplification, convenient maintenance, highly
automation and minimal operators to obtain enhanced safety and improved economic efficiency.

4. Main Design Features
Nuclear Steam Supply System
Not Applicable (N/A).

Reactor Core
The core of DHR400 consists of 69 fuel assemblies. Each fuel assembly is 2.15m long and its design is
modified from the standard 17×17 PWR fuel assembly with 264 fuel rods, and 9 supporting tubes. The fuelis UO2 with Gd as a burnable absorber. The 235U enrichment is below 5%. The reactor operates 150 days per
year (typical 5-month winter period in northern China) and a three-batch refuelling is conducted off power
on a 15-month refuelling cycle, that is, the reactor is refuelled every 3 years.

Reactivity Control
Reactivity control during normal operations is achieved through control rods. Two separate shutdown
systems are used for reactor shutdown and scram events. Fuel rods containing different mass ratios ofgadolinium oxide are introduced in all new fuel assemblies to maintain reactivity during burnup and to
flatten the axial power profile.

Reactor Pool and Internals
The reactor pool is a cylinder with an inside diameter of 10 m and an overall height of 28 m, containing thereactor core and in-pool structures inside its 25 m depth of water. The pool is buried underground with anelevation of its bottom at -27m. The pool is made of reinforced concrete with an inner layer of duplex
stainless steel. The upper head includes a carbon steel truss and a stainless-steel plate, connected to theconcrete wall of the pool, and provides support for the control rod driven mechanism and the control rod
guide tubes. Below the upper head there is a plenum, which is connected to an engineered venting system toexhaust vapour and other gases. Internal structures are designed to support the core and isolate the water
inside the core from the water in the pool. The large water inventory in the pool provides large thermal
inertia and a long response time, thus enhances the resistance to system transients and accidents. Thesefeatures ensure that the core will not melt down under any accident.

Primary Coolant System
During normal operations, the 68oC inlet water enters the core from the bottom, the water is then heated to
98oC. The outlet water enters the 4 outlet pipes through the rising barrel right above the core and then enters
4 separate primary pump room through the pool wall. The water entering the primary pump room flows into
the primary heat exchanger. Inside the primary heat exchanger, the primary water is cooled to 68oC by thesecondary side water and then pumped back to the upper part of the reactor pool. The water then flows down
to the bottom and enters the core again. No boiling will occur during normal operations.

Secondary Coolant System
The secondary coolant system is an isolated sealed intermediate loop that separates the primary loop and thetertiary loop while transfer heat from the primary to the tertiary. The pressure of the secondary loop is
designed to be higher than that of the primary loop, so there is no chance that the radioactive primary water
will contaminate the tertiary loop.

Primary Heat Exchanger
The DHR400 uses 8 plate heat exchangers in its primary loop to transfer heat to the secondary loop. Plate
heat exchanger is suitable for low temperature difference water-to-water heat exchange for its small
resistance and high efficiency. The leak tightness of the plate heat exchanger is highly reliable. Even under
the circumstances of leakage, the coolant leaks outwards to the pump room. This feature provides advantages
to radioactivity isolation.

5. Safety Features
The DHR400 is designed with inherent safety features. These include large water inventory in the reactor
pool, two sets of reactor shutdown systems, pool water cooling system and a decay heat removal system with
passive heat pipes. With these designs stable long-term core cooling under all conditions can be achieved.

Engineered Safety System Approach and Configuration
Instead of augmenting additional engineered safety systems the DHR400 emphasise on inherent safety
features. The great heat capacity of the 1800 tons of water inside the reactor pool ensures that the reactor
28

core will be kept submerged in all circumstances, thus no core melt down could occur. It has negative
temperature and void reactivity feedback; therefore the power increase can be effectively restrained. In theevent of severe accidents, the reactor can automatically shutdown by the inherent negative reactivity
feedback, and the reactor core will be kept submerged for quite a long time even with no further intervention.

Decay Heat Removal System
In normal shutdown situation, decay heat is removed by primary coolant system. When the primary pump is
shutdown, the decay heat is removed from the reactor core to the reactor pool with natural circulation. And
the heat of the pool water is transferred to the environment with pool water cooling system and passive heatpipe system.

Emergency core cooling systemNo emergency core cooling system is employed in the DHR system. The reactor pool with 1800 tons ofwater provides an immense heat sink. Even under the circumstances that all cooling system failed, corecooling can be maintained with the evaporation of the pool water for several days. So, emergency corecooling system is not necessary for DHR.

Spent fuel cooling safety Approach/System
Spent fuel pool cooling and purification system is designed to remove the decay heat released from the spent
fuel and maintain the pool water temperature within an acceptable limit.

Confinement System
There are three barriers precluding a radioactive release to the environment in DHR, including the fuelcoating, the reactor pool and the confinement building on top of the pool. Due to the low operating
temperature and atmospheric pressure above the reactor pool, there are no high-pressure events and insteadof a containment, a confinement building is sufficient for protection. The reactor is located underground andthe reactor core is submerged in 1800 tons of water, making DHR400 highly resistant to external events.
Additional protection is provided by the reactor confinement building above the pool.

Waste Gas Treatment System
A waste gas treatment system is adopted to store and decay short life gaseous fission products.

6. Plant Safety and Operational Performances
Two independent systems are provided for reactor power control and to ensure safe reactor shutdown.
Reactor cold start-up and rapid start-up can be achieved safely due to the negative temperature reactivity
coefficient.

7. Instrumentation and Control System
The instrument and control (I&C) system of the reactor will adopt distributed instrument control system toprovide consistent and effective human-machine interface, improve the availability, reliability, operability,and human-machine interface design of the system, and reduce the requirements for spare parts with aconsiderable design flexibility.

8. Plant Layout Arrangement
The layout of the DHR400 is illustrated below. The
main building is arranged in a single unit, which
combines the reactor confinement building, auxiliarybuilding, and exhaust tower. The reactor confinementbuilding is the core of the main building, and theauxiliary building includes the heat supply building,
the radioactive waste treatment and storage building
and other auxiliary buildings. It is arranged aroundthe reactor building to facilitate the connection
between the buildings.

Reactor Confinement Building
The reactor confinement building includes reactor
pool, spent fuel pool, storage pool, reactor hall andprimary loop system room.

Heating Building

Plant Layout

The heat supply building includes the water pump room and the secondary heat exchanger room, which arelocated at one side of the reactor confinement building and transfer the core heat from the primary loop to theheat supply network.
29

Radioactive waste treatment and storage Building
The radioactive waste treatment and storage building is located on the other side of the reactor confinementbuilding. It collects and stores radioactive liquid, solid and gaseous wastes generated during operation andmaintenance to meet the requirements of radioactive waste management.

Other auxiliary buildings
Other auxiliary buildings include new fuel room, circuit purification system room, spent fuel pool coolingsystem room, equipment cooling water room, supply and exhaust fan room, electrical equipment room, main
control room, electronic equipment room, physical protection centre, etc.

Exhaust tower
The exhaust tower is located at the exhaust fan room on the 5.000m floor, with a total height of 40m. Fromthe elevation of 11.000m to 24.000m, it is a reinforced concrete structure, and from the elevation of 24.000m
to 40.000m, it is a steel circular exhaust duct. The radioactive gas (radioactive aerosol and radioactive
iodine) from the radioactive area of the main building is first filtered and then discharged at high altitude
through the exhaust tower to meet the environmental radiation safety requirements.

9. Design and Licensing Status
The basic design of DHR has just completed (some parameters might change with the optimization of
DHR400 design). Seeking for construction license in 2023.

10. Waste Management and Disposal Plan
The radioactive waste management system is designed for the storage and treatment of radioactive waste.
The system includes radioactive waste liquid collection and treatment system, radioactive waste resin
collection and treatment system, solid waste collection system and radioactive waste gas collection andtreatment system.

11. Development Milestones
2023
2026

30

Construction licence
Commercial operation

HAPPY200 (SPIC, China)
All content included in this section has been provided by, and is reproduced with permission of the SPIC, China.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type

State Power Investment
Corporation, Ltd. (SPIC), China
PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

200 / 0 (thermal power only)

NSSS Operating Pressure(primary/heating), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

0.6 / 0.8

Number of fuel assemblies in
the core
Fuel enrichment (%)

37

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

40

Reactivity control mechanism

Rods

Approach to safety systems

Active / Passive

Design life (years)

60

Plant footprint (m2)1150

RPV height/diameter (m)6.0 / 2.25

RPV weight (metric ton)

26

Seismic Design (SSE)

0.3g

Fuel cycle requirements /ApproachDistinguishing features

LEU / once through

Design status

Forced (2 pumps)

80 / 120
UO2 / Square 17x172.76 avg / 4.45 max

18

Pool-loop combined typereactor, heat generator.
Detailed Design

1. Introduction
The Heating-reactor of Advanced low-Pressurized and Passive SafetY system – 200 MW(t) (HAPPY200) is aso called pool-loop combined type reactor, which has both features of swimming pool reactor and PWR tosome extent. The HAPPY200 operates under low temperature and at low pressure in the closed primary circuit
boundary with forced circulation mode. All engineered safety systems operate in a passive mode, and can copeat least for 1 month after accident without any active operator intervention, taking advantage of external cold
air as the ultimate heat sink. A large pool of water inventory is incorporated to enhance its safety and reliability
under postulated accident conditions. The key design objectives of HAPPY200 include: inherent safety; good
economy; proven technology; and easy decommissioning. Aiming for high reliability with inherent safety
features, HAPPY200 can be deployed in the vicinity of the targeted heating supply district or community withhigh population density.

2. Target Application
The HAPPY200 is designed for heat generating power source, dedicated to provide northern cities in China
with a clean heating solution, and can be operated for 4 – 8 months each year during the winter. Its non-electric
applications include sea water desalination, house cooling in summer, energy storage etc. without any need ofsubstantial design change.

3. Design Philosophy
The HAPPY200 draws on the mature operating experience of light water-cooled reactors, pool reactors and31

passive nuclear safety technology. Both RCS and ESF system are operated under low pressure and low
temperature spectrum, and the design parameters can be fine-tuned to adapt to local civil heating requirement.
HAPPY200 adopts proven technology and equipment, such as truncated fuel assembly, plate-type heatexchanger, etc. These equipments have full operational record, high reliability and high maintainability. The
plant is to be built half-underground, with reactor core vessel totally submerged in a large underground pool,to elimination of external-event challenge. And 5 FP shielding and barriers are deployed to ensuring
radioactivity isolation from terminal users, as well as the surrounding public. The primary safety objective of
HAPPY200 design is practical elimination of core melting and technical cancellation of off-site emergency,
so that the reactor can meet basic evaluation principle for heating reactor formulated by the Chinese National
Nuclear Safety Administration.

4. Main Design Features
Primary Circuit
During normal operation, the 80℃ inlet water enters into the corefrom the bottom, and the water is then heated to a temperature of
120℃. The outlet water enters the 2 hot legs and then enters four
separate primary heat exchangers. Inside the primary heatexchanger, the primary water is cooled to 80℃ by the secondaryside water. The water flowing through the primary exchangers
enters 2 cold legs separately. The water then flows down to thebottom and enters the core again. No boiling will occur during
normal operations.

Reactor Core and Fuel
The core of the HAPPY200 consists of 37 fuel assemblies. Eachfuel assembly is 2.1 m long and its design is modified from astandard 17×17 PWR fuel assembly with 264 fuel rods. The fuelis UO2 with Gd as a burnable absorber. The 235U enrichment isbelow 5%. The reactor operates 180 days per year (typical 6month winter period in northern China) and a three-batchrefuelling is conducted off power on an 18-month refuelling
cycle.

Secondary SideThe secondary coolant system is an isolated sealed intermediate
loop that separates the primary loop and the third loop whiletransfer heat from the primary to the third. The pressure of thesecondary loop is designed to be higher than that of the primaryloop, so there is no chance that the radioactive primary side water
could contaminate the third loop.

Reactivity Control
HAPPY200 uses 21 control rod clusters to provide enoughreactivity compensating capacity, with magnetic force type andhydraulic force type control rod driving mechanism (CRDM). No chemical shim (e.g. Boron) is used forreactivity control. HAPPY200 blends a lot of gadolinium oxide in fuel rods. Because there is non-solubleboron in the core, the reactor is operated by shifting control rods to maintain criticality.

Reactor Pressure Vessel and Internals
The reactor pressure vessel is submerged inside the large water pool, isolated from the pool during normal
operation and other conditions, RCS is connected to pool only if the RCS is depressurized and need injectionof cooling water from the pool.

5. Safety Features
The primary safety objective of HAPPY200 design is practical elimination of core melting and technicalcancellation of off-site emergency. To achieve this safety objective, the safety concept of HAPPY200 is basedon inherent safety features, the defence in depth principle, the use of passive systems to prevent accidents andmitigate their consequences, and multi-barriers to the release of radioactive materials into the environment.

Engineered Safety System Approach and Configuration
HAPPY200 has many characteristics including low temperature, low pressure parameters, high thermal safetymargin, negative power reactivity, simplified engineered safety features (ESF), etc. And the system uses
passive cooling system and anti-seismic system. The safety systems of HAPPY200 consist of: redundant
shutdown system, passive feed-bleed system (PFB), passive residual heat removal system (PHR), passive pool
air cooling system (PAC), etc. These systems could maintain core integrity through the plant lifetime.
32Decay Heat Removal System
The passive residual heat removal system is designed toremove the residual heat from the reactor core during
unexpected operational occurrences and events, especiallyfor the non-LOCA events. It consists of two series verticaltube heat exchangers connected with the hot legs and cold
legs of each loop. The passive residual heat removal modeis provided by the natural circulation of the coolant inprimary circuit through the heat exchangers, which use thelarge capacity pool water as the heat sink. The heat sinkcapacity is adequate to passively cool down the reactor andprevent hazardous superheating of the core.

Emergency Core Cooling System
Pool water heat sink is used for emergency core coolant
injection and residual heat removal.

Containment System
The containment of HAPPY200 is a steel shell type, acting
as the last barrier of fission product. It also partiallyfunctioned as heat removal sink to environment during
normal operation and accident condition.it is partiallydeployed underground. A total of 5 safety barriers: fuelpellet, fuel cladding, primary circuit system, large volume
shielding pool and steel containment.

Engineered safety features:-1 Passive Residual Heat Removal System-2 Passive Safety Injection System-3 Passive Air Cooling System

6. Plant Safety and Operational Performances
HAPPY200 unit itself does not consider generating electricity. In normal operation, external power supply isrequired to ensure the operation of each device and system. In accident conditions, the unit can be returned tothe shutdown state and take away the decay heat without relying on the external power supply by the inherent
safety features and passive safety system.

7. Instrumentation and Control Systems
The instrumentation and control (I&C) system provides the capability to monitor, control and operate plantsystems. The I&C system is implemented using mature technology based on high economy.

8. Plant Layout Arrangement
HAPPY200 is designed to be sited in inland areas near the targeted heating supply area centre. No special
requirements to adapt with the local air temperature, humidity and other conditions. The principal structures
are the reactor building, electricalbuilding, fuel building and auxiliarybuilding. HAPPY200 unit itself does
not generate electricity. In normal
operation, external power supply isrequired to ensure the operation ofeach device and system. In accident
conditions, the unit can be returned tothe shutdown state and take away the
decay heat without relying on theexternal power supply by the inherent
safety features and passive safetysystem. HAPPY200 does notdischarge waste heat or require alarge amount of cooling water,
because the plant design adopts
closed cycle. By using closed cycle,
the amount of water replenished innormal operation is small and all of the heat is supplied to the heat consumer. HAPPY200 with smaller or zero
emergency planning zones (EPZ) up to only 3 km in radius is expected to be approved by the regulators.

Engineered Safety System Approach and Configuration
The reactor building is used to prevent the radioactive materials escaping to the environment at the conditionof LOCA accident. At normal conditions and accident conditions, it provides radiation protection and protects
the internal systems from external disasters. The reactor building is mainly used to arrange reactors and otherprimary loop equipment, such as main pump, primary/secondary heat exchanger, pressurizer, shielding pool,33

timeless air-cooling system, chemical and volume control system, equipment cooling water system, NI
ventilation and air conditioning system, etc.

Electrical building
The electrical building is mainly used to arrange power distribution equipment, instrumentation and control
equipment, main control room, battery, ventilation system, fire protection system, and so forth.

Fuel building
The fuel building is mainly used for equipment layout and operation of fuel handling, transportation andstorage systems. It is also used to arrange pool cooling and purification system, ventilation and air conditioning
system, chemical and volume control system, etc.

Auxiliary building
The liquid waste treatment system is performed at the auxiliary building. The system includes the radioactive
liquid waste recovery system, nuclear sampling system, exhaust gas treatment system, and nuclear auxiliarybuilding ventilation system.

9. Testing Conducted for Design Verification and Validation
The HAPPY200 reactor system is in general designed based on proven technology and method. The spectrumof reactor core and coolant temperature and pressure is decreased consequently both under the normal
operation and accident conditions, which resulted in larger thermal margin compared to conventional LWR
design criteria, as well as possible cooling and flooding of reactor core accessible to low pressure system, as
part of ESF features. To date, 3 important test facilities, one for secondary reactor shutdown system validation
testing and one for engineering design validation of PFB and PHR system, and in addition, a large steel
containment for validation of PAC system has been built and ready for performing of validation test. Besides,
a test for critical heat flux of truncated fuel assembly is also planed for near-term construction.

10. Design and Licensing Status
HAPPY200 has completed conceptual design. Preliminary design is underway. The commercial demonstration
project is carrying out preliminary work of the project, and the site selection and preliminary feasibility analysis
have been completed. HAPPY200 meets regulatory requirements for design and licensing in most countries.
The first project has completed the site selection and preliminary feasibility analysis report review. The next
step will submit site safety assessment report and site stage environmental impact assessment report to theChinese regulatory authorities

11. Fuel Cycle Approach
Because in the north of China, heating system is required to work about six months in a year, the HAPPY200
reactor should be operated about 18 months in the 3-year cycle, the reactor should be shutdown for six months
every year. However, there is no need for refuelling in one cycle for 3 years. The enrichment of uranium used
by HAPPY200 is less than 5%, and the average discharge burnup of fuel assemblies is about 40 GWd/tU.

12. Waste Management and Disposal Plan
After spent fuel discharge from the reactor, it is stored on plant site and then cooled and decrease inradioactivity for some years. Once-through fuel cycle option is considered after plant site decommissioning.
Spent fuel leave at safety level and transfer to specific disposal site.

13. Development Milestones
2015
2016
2019
2025
2027

34

Start market investigation and concept design (changes)
Concept design completed
Start pre-project work in north of China
To start construction of a proto-type NPP in China
Target commissioning and commercial operation dates

NHR200-II (Tsinghua University and CGN,
China)All content included in this section has been provided by, and is reproduced with permission of, Tsinghua University/CGN, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
Tsinghua University and CGN,of origin
China
Reactor type
Coolant/moderator

Integral PWR
Light Water / Light Water

Thermal/electrical capacity,
MW(t)/MW(e)200 MW(t)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Natural Circulation
8

Core inlet/outlet coolant
temperature (oC)

232 / 280

Fuel type/assembly array
Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)UO2 pellet / square 9 × 9
208

Reactivity control mechanism

The hydraulic control rod drive
mechanisms
Passive60–
13.5 / 4.94 (max, external
diameter)388
0.3 g
in-out refuelling

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

1.8, 2.67, 3.4
24
30.8 (average)

Distinguishing features

NaturalSafetyDesign status

Basic design

Circulation,

Passive

1. Introduction
NHR200-II is an integrated light water reactor adopting full-power natural circulation in the primary loop andpassive safety features which is developed by Tsinghua University and China General Nuclear Group (CGN).
NHR200-II is based on the 5 MWt experimental Nuclear Heating Reactor (NHR) and 200 MWt Nuclear
Heating Reactor (NHR200-I) which are designed by Tsinghua University.

2. Target Application
NHR200-II is a multipurpose reactor. It can be used for district heating, industrial steam supply, steam/watercogeneration, heat-electricity cogeneration and seawater desalination.

3. Design Philosophy- Integrated reactor- Full-power natural circulation- Passive safety feature and simplification systems- Near to the city or customers

35

4. Main Design Features
Nuclear Steam Supply System
The integrated nuclear steam supply system (NSSS) design consists of the reactor core and fourteen primaryheat exchangers. The heat of core is transferred to Steam Generator (SG) by the intermediate loop.

Simplified diagram of NSSS.

The simplified diagram of NSSS is shown as above.

Reactor Core
The core of NHR200-II consists of 208 fuel assemblies and 52 control rod assemblies. Each fuel assembly
consists of a fuel bundle and a Zircaloy-4 channel which surrounds the fuel bundle. The fuel is UO2 with Gd2O3
as a burnable absorber homogeneously mixed within the fuel for select rod locations. The control rod
assemblies are inserted into the gap between four fuel assemblies. The reactor has a refueling cycle of 24months.

Reactivity Control
Reactivity control is achieved through 52 control rod assemblies. The control rod assembly consists of asheathed cruciform array of stainless steel tubes filled with boron carbide powder. The hydraulic control rod
drive mechanisms (CRDMs), which are designed by INET, are placed in the reactor pressure vessel (RPV)
and can eliminate the possibility of postulated rod ejection accidents.

Reactor Pressure Vessel and Internals
The RPV consists of a cylindrical steel vessel with an inside diameter of approximately 4.3m and an overall
height of approximately 13.5m. The main pipes are eliminated and small-size nozzles are placed on the higher
part of the vessel, or on the upper head, so the feasibility of a primary coolant loss accident would be much
lower. And the lower part of the vessel has two layers. For the most serious condition, a break occurring on
the bottom of the RPV, the coolant will be contained by the second layer of the vessel, assuring that the reactor
core can always be covered by water. 14 primary heat exchangers are placed between the core barrel and RPV,
vertically higher than the core.

Reactor Coolant System
The NHR200-II core, which is located in the lower part of the RPV, is cooled through natural circulation
without reactor coolant pumps. When flowing through the reactor core, the coolant is heated and flows upward
into the riser. After flowing through the primary heat exchangers, where it is cooled by the water of theintermediate loop, the core coolant increases in density, naturally flows downward and returns to the core inlet.
The pressure of the reactor coolant system is automatically maintained by the nitrogen/steam space in the topof the RPV.

Steam Generator
Each NHR200-II has 14 primary heat exchangers located in the RPV and 2 steam generators located outside
the containment. Two intermediate loops, each of which connects one steam generator and seven primary heatexchangers, are deployed to carry heat from the reactor and transfer it to the water/steam of the second loop.
The primary heat exchanger is tube-in-tube type and the core coolant flows inside the inner tubes and outside
the outer tubes, while the intermediate loop water flows in the gaps between the inner and outer tubes.
36

Pressurizer
Not Applicable. No separate pressurizer is existed in NHR200-II. And the pressure of the reactor coolant
system is automatically maintained by the nitrogen/steam space in the top of the RPV.

Primary pumps
The NHR200-II is a kind of natural circulation reactor.

5. Safety Features
Engineered Safety System Approach and Configuration
As NHR200-II has been designed with a number of advanced and innovative features, only a few engineered
safety systems are needed, including the decay heat removal system, the passive boric acid injection system,and the containment system. The passive safety systems are shown below.

Passive Engineered Safety Systems

Decay Heat Removal System
Two trains of passive decay heat removal loops are deployed independently. Each train contains of 7 primaryheat exchangers (second side), a set of air coolers and associated pipes and valves. The core decay heat afterreactor shutdown is finally dispersed into the atmosphere by natural circulation. There is no need of a power
supply, so the decay heat removal will be ensured. The core decay heat can also be dispersed into the plant
service water system via the heat exchangers of the chemical and volume control system, which is not a safetyclass system.

Emergency Core Cooling System
Not Applicable.

Containment System
The reactor and some auxiliary systems are placed in a small concrete containment. According to our design,
the containment heat removal system is not needed. At the same time, a higher leak rate is permitted because
the accident radioactive source term is much lower than that of the nuclear power plants in operation.

6. Plant Safety and Operational Performances
According to the basic design, a preliminary PSA of internal events of NHR200-II has been carried out, theresult of which shows that the core damage frequency (CDF) is less than 1E-8/reactor-year. Because thesystems are highly simplified, the operation of NHR200-II is so easy that only 5 operators are needed in a tworeactor-module control room.

7. Instrumentation and Control System
The instrumentation and control system of NHR200-II is similar to those of normal PWR plant. The two reactor
modules are controlled in a coordinated manner to meet various operational requirements.

8. Plant Layout Arrangement
The nuclear island contains the reactor building and nuclear auxiliary buildings. The spent fuel storage pool ishoused in the containment, so the extra fuel storage building is eliminated. For different applications, different
equipment can be arranged in the conventional island, such as steam turbo-generators, heat exchanger stations,or seawater desalination equipment.

37

Nuclear island and conventional island layout

9. Design and Licensing Status
The basic design of NHR200-II has been completed and the preliminary safety analysis report (PSAR) is
published. In Jan. 2018, based on the preliminary feasibility study results, the feasibility study of thedemonstration nuclear heat-supply plant in Hebei Province was permitted by China National Energy
Administration.

10. Waste Management and Disposal Plan
Waste management approach and disposal plan is similar to other nuclear power plants, and different waste
management and disposal technology could be supplied for different site.

11. Development Milestones
1989
1996
2006
2016
2022

38

5MWt experimental reactor was constructed
The design of NHR200-I was completed
INET began to develop NHR200-II
Basic design of NHR200-II was completedPlan to complete the preliminary work of NHR200-II demonstration project

TEPLATORTM (UWB Pilsen & CIIRC CTU,
Czech Republic)
All content included in this section has been provided by and is reproduced with permission of UWB Pilsen & CIIRC CTU Prague, Czech Republic.

MAJOR TECHNICAL PARAMETERSValue
UWB Pilsen & CIIRC CTU
Technology developer, country ofPrague, Czech Republic
originParameter

Reactor type

Channels in Reactor Vessel

Coolant/moderator

Heavy Water (D2O)/ Heavy
Water (D2O)<150 / does not produce
electricity
Forced circulation

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (°C)
Fuel type/assembly array
Number of fuel assemblies
Fuel enrichment (%)
Core Discharge Burnup (GWd/ton)Refuelling Cycle (months)
Reactivity control mechanism<2/<2< 175 / < 192
VVER-440 / hexagonal with
126 fuel pins
55
Spent fuel (< 1.2 wt% 235U
equivalent) or slightly enrichedfresh fuel2.3
10 months with the online
optionControl rods, Moderator height

Design life (years)

Inherent and passive safety withbuilt/in decay heat sink
60

Plant footprint (m2)

≤ 2 000

RPV height/diameter (m)

8/4
N/A

Approach to safety systems

RPV weight (metric ton)
Seismic Design (SSE)
Fuel cycle requirements /ApproachDistinguishing features

Design status

0.2 g
LEU - reuse of LWR spent FAs,
possibility to run on fresh SEU
District heating zero CO2
source with zero fuel cost, heatstorage for decay heat removal.
Conceptual design

1. Introduction
The TEPLATOR is an innovative concept for future district and process heat production or chilling
applications. The TEPLATOR facility will use already irradiated fuel from conventional light waters reactor(which are not burnt up to its regulatory and design limits). In order to harvest additional energy from alreadyused FAs, the TEPLATOR is a nuclear reactor derived by state-of-the-art computational tools using better
moderation, more optimal fuel lattice pitch, lower fuel temperature and lower coolant pressure. Different
TEPLATOR variants are proposed, using either used BWR, PWR or VVER irradiated fuel assemblies or a
fresh, slightly enriched fuel with an output power range between 50 and 200 MW(t). The TEPLATOR DEMO
will operate at 50 MW(t) output power with fresh VVER440 fuel.

2. Target Application
The TEPLATOR is designed for clean district heating energy production for cities with 100 000 or more
inhabitants. It will replace the conventional heating plants based on fossil fuels. The TEPLATOR will produce
heat without any emissions or carbon footprint and with negligible fuel costs. TEPLATOR solutions areespecially suitable for countries with thousands of LWR FAs stored either in interim storage casks or spent
fuel pools. These FAs are now a financial liability that, once used for heat production, can turn into a sizeable39

financial asset. The calculated investments cost for the first TEPLATOR DEMO 50 MW(t) facility is 30 M
EUR. Then the final price of produced heat is 4 EUR/GJ (using input prices for 2019).

3. Main Design Features
Design Philosophy
The design philosophy uses only proven, known, verified and tested high TRL components. This ensures low
investment costs and low risks. The design itself includes three circuits. The primary circuit consists of a socalled calandria, a core with the spent LWR fuel FAs, three heat exchangers and three pumps. The core ismade from Zr channels in which the fuel is inserted. Heavy water as a moderator fills the space between thechannels. The coolant flows in all channels through a system of pipes to the collector. Three pipes are led outof this collector, each of which is led into a separate heat exchanger. The coolant passes through the primaryside of the heat exchanger and returns to the fuel channels through the pump and the lower distribution
chamber. A secondary (intermediate) circuit transfers the heat from the primary circuit to the district heating
circuit. The secondary circuit heat transfer fluid (HTF) could either be water or another fluid (based on theoperating parameters). The secondary circuit includes two storage tanks connected to the circuit serving as anenergy storage system for shaving off demand peaks. These storage tanks can simultaneously dissipate andstore heat from the residual power of the fuel, i.e. the intermediate tanks are designed to absorb decay heat of
the core in DBAs. The tertiary or district heating circuit, which distributes the heat to the end customer, isseparated from the core by two heat exchangers.

Nuclear Steam Supply System
The TEPLATOR operates only in the liquid phase, so no steam generation or electricity production is expected.

Reactor CoreThe TEPLATOR core consists of equally spaced channels filled with spent nuclear fuel from LWR reactors.
More customisation options are possible. The initial one is the reuse of VVER-440 spent nuclear fuel. In thatcase, totally 55 fuel assemblies are placed in a large-pitched hexagonal array. Typical VVER-440 spent nuclear
fuel uses 3.6 wt% 235U initial enrichment, 35 GWd/ton average burnup and 30 years of cooling. Alternativeuse of slightly enriched fresh fuel is possible. Each fuel assembly is placed in a coolant tube filled with heavy
water or alternatives for temperatures up to 192 oC. The atmospheric pressure of a heavy water moderator
eliminates the need for a thick and expensive pressure vessel, the temperature of the moderator will be lower
than 60 °C. The TEPLATOR is a heat generator with a typical operation time of 10 months each year (regular
European heating season).

Reactivity Control
Two independent systems are deployed. Reactivity control under normal operation is achieved by changing
control rods positions. The second reactivity control system is based on moderator pumping in or draining out
the reactor calandria. The safety shut-down system is based on a combination of the control rods and moderator
draining systems.

Calandria and Internals
The internals of TEPLATOR consist of the fuel channels, channel outlets, control rods, control rods drive
mechanism, I&C systems, axial and radial reflector, and bottom collector. The coolant (heavy water) is
distributed through the bottom collector to individual channels. The calandria works on ambient pressure, itdoes not need to be pressurized. The space between fuel channels and calandria is filled with heavy water,which serves as a moderator. The total volume of heavy water in calandria is around 30 m3. A graphite reflector
surrounds the core both in the axial and radial directions.

Reactor Coolant System
The primary coolant (D2O) enters the core of TEPLATOR at the temperature of 175 °C. It flows through thefuel channel, and then it leaves the individual channels at 192 °C at the channel outlet. This outlet is attachedto the collector where the primary coolant is collected. From the collector, the coolant is distributed to the threeheat exchangers where it heats the secondary heat transfer fluid (HTF). The primary coolant flows through thepump, then through the pipe on the inside of the calandria to the bottom collector where it is distributed again
to the individual channels. Roughly 20 m3 of D2O is required in the primary circuit.

Primary Heat Exchanger
The TEPLATOR is a three-loop design. Thus it has three primary heat exchangers (HE) to transfer the heatfrom the primary to the secondary circuit. The heat exchanger is a horizontal type with U-shaped tubes andwater-water heat exchange. Each of the HE has a heat transfer surface about 520 m2 and is capable, underforced circulation, of cooling the TEPLATOR core on its own: decay heat under emergency conditions can besafely removed by HE to the energy storage tanks using natural circulation.

Pressurizer
The pressurizer is linked to the heavy water management systems, and it operates at low pressure (2 MPa).
40

Secondary side
The secondary circuit is an intermediate loop that separates the primary circuit and the tertiary circuit whiletransferring heat from the primary to the tertiary circuit. The secondary circuit consists of the secondary sideof primary heat exchangers (HE I) and the primary side of the secondary heat exchangers (HE II). As part ofthis circuit the energy storage system, consisting of two tanks, can be connected having identical heat transfer
fluid (HTF) as the secondary HTF. This energy storage system is based on thermal energy storage (TES) heatmechanism which serves several purposes: 1) shaving off TEPLATOR power fluctuations, 2) compensationand smoothing of the demand curve and 3) emergency and safety heat sink.

4. Safety Features
General safety features
The TEPLATOR operating conditions (e.g., fuel/coolant temperature, pressure, linear heat production) aremuch lower than those for which the used FAs were certified and used in LWRs. The safety features establish
defence-in-depth against radiological hazards. The TEPLATOR leverages the inherent safety characteristicsof the basic LWR reactor design and supplements them with passive and active safety features that emphasizes
proven and result in an improvement in safety. The TEPLATOR secondary circuit provides large volumes of
fluid that are available to provide cooling to the core in the event of accidents, including by passive means.
The TEPLATOR places all reactivity devices in low-temperature, low-pressure moderator, eliminatingpressure-driven ejection of reactivity devices from the core. The separation of moderator from coolant also
provides two separate heat removal means in the event of accidents and ensures that moderator temperature
feedback to the core physics is negligible in normal operation.

Engineered Safety System Approach and Configuration
The TEPLATOR has two separate shutdown systems. These are two fully-capable fast-acting means of
shutdown for use at the third level of defence in depth, fully independent of each other.

Decay Heat Removal System and Emergency Core Cooling System
Decay heat removal system is integrated as the energy storage system interconnected to the secondary circuit.
During TEPLATOR shut-down, heat generated in the core is transported by natural circulation inside thecooling loops. This heat is removed in the primary heat exchanger using thermal energy storage (TES). The
TES system consists of two tanks, a ‘cold' and a 'hot’ one. In order to remove decay heat from the TEPLATOR,
heat transfer fluid (HTF) from the cold tank flows via natural convection through the secondary side of the
primary heat exchanger (HE I) to the hot tank. The volume of both tanks is designed to be sufficient forremoving decay heat for long enough that the auxiliary cooler dissipates the heat.

Containment System
The TEPLATOR containment system includes a reinforced concrete structure (the reactor building) with areinforced concrete dome and an internal steel liner, access airlocks, equipment hatch, building air coolers forpressure reduction, and a containment isolation system.

5. Plant Safety and Operational Performances
Two independent systems are provided for reactor power control and to ensure safe reactor shutdown. Reactor
cold start-up and rapid start-up can be achieved safely due to the large negative temperature reactivity
coefficient.

6. Instrumentation and Control Systems
The instrumentation and control (I&C) system provides the capability to monitor, control and operate plantsystems. The I&C system is implemented using mature technology; i.e., known and tested LWR detectors andsystems.

7. Plant Layout Arrangement
The plan layout of the TEPLATOR is
illustrated below.

Reactor Building
The TEPLATOR facility consists ofone main structure which further
contains nuclear and non-nuclear
sectors/buildings. Nuclear sectors are
the main TEPLATOR hall, the fuelhandling building, and the auxiliary
nuclear building. Non-nuclear sectors
are the heat exchanger hall and theauxiliary building. Other buildings andstructures within the facility layout are41

a heat accumulator, water
storage tanks, auxiliary cooling
towers with a pumping station,transformers of power supply
and backup diesel generators.

Balance of Plant
Heating / Chilling Supply
system is located in the heatingexchanger hall next to the main
TEPLATOR hall. An option to
use the TEPLATOR heatsource for district chilling
solutions is also investigated.

8. Design and Licensing
StatusThe TEPLATOR project completed its preconceptual design and the works on basic design started in the Q4of 2020. The commercial demonstration unit siting negotiations– TEPLATOR DEMO with thermal power of
50 MWt is in the preliminary phase. The preliminary phase includes the feasibility study, the site location
selection and obtaining the license for construction. Once the feasibility study is done and the site location isapproved, the environmental impact assessment report will be carried out and will be submitted to theregulatory authorities.

9. Fuel Cycle Approach
The unique feature of TEPLATOR is the reuse of spent nuclear fuel from commercial LWRs which is normallyconsidered a waste. This can be achieved due to significantly lower operation parameters (well under limits)
when compared to the conditions in large LWRs. Based on heating or cooling demand, the core can be operated
from 10 months each year with subsequent refuelling of fuel assemblies. Usage of fresh SEU assemblies ispossible and increases the heat price by roughly 20%.

10. Waste Management and Disposal Plan
When removed from the core, reused fuel will be stored and cooled in the fuel handling building and thereafter
transported back to the original spent fuel storage or as required by the national law.

11. Development Milestones
2019- 2020
2020-2021
2021-2024
2024-

42

Preliminary studies and pre-conceptual design
Conceptual design phase and technology validation
Design phase - Basic design
Detail design / Licensing
Start of DEMO unit construction, first unit in service in 2028+

NUWARDTM (EDF, France)
All content included in this section has been provided by, and is reproduced with permission of EDF, France.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
country of origin

Reactor type

EDF, France with major
contributions from CEA,
Naval Group, Framatome,
TechnicAtome, andTractebel-Engie
Integral PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

2 x 540 / 2 x 170

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

15 / 4.5

Number of fuel assemblies
in the core
Fuel enrichment (%)

Forced circulation

280 / 307
UO2 / 17x17 square pitch
arrangement
76<5

Refuelling Cycle (months)

24 (half core)

Reactivity control

Approach to safety systems

Control rod drive
mechanism (CRDM),
solid burnable poisons
Passive

Design life (years)

60

Plant footprint (m2)
RPV height/diameter (m)3500, nuclear island
including fuel storage pool
15 / 5

RPV weight (metric ton)

310

Seismic Design (SSE)

0.3g

Distinguishing features

Integrated NSSS with poolsubmerged containment,boron-free in normal
operation and in all
Design Basis Conditions(DBC), semi-buried
nuclear island
Conceptual Design

Design status

1. Introduction
NUWARDTM is an integrated PWR design to generate 340 MWe from two independent reactor units, offering
flexible operation. The design incorporates the main components of the Nuclear Steam Supply System (NSSS)
including Control Rod Drive Mechanism (CRDM), Compact plate Steam Generators (CSGs) and pressuriser,
all contained within the Reactor Pressure Vessel (RPV). The RPV is then installed within a further layer ofsteel containment and this structure is immersed in a pool filled with water (the reactor pool). The NUWARDTM
design includes the capability to cope with Design Basis Conditions (DBC) using passive systems without theneed for any external electrical power supply. The reactor is self-reliant, connected to an internal ultimate heatsink (the reactor pool) which offers a coping time of more than 3 days without the need for intervention.

43

2. Target Application
The NUWARDTM technology is being developed to replace fossil-fired power plants in the 300-400 MWe
range; to supply power to remote municipalities and energy-intensive industrial sites; and to power grids withlimited capacity. By design, it is a multipurpose SMR that can be used for cogeneration of heat and electricity,hydrogen production, district heating, and water desalination. The design offers baseload and load-following
capability to enable integration with renewable energy sources.

3. Design Philosophy
NUWARDTM design is based on the proven PWR-technology that incorporates significant experience acquiredin the fields of medium and high-power generation, alongside several key technological innovations, to achieve
the following design objectives:(i)
Acceptability: robustness of the design will maximise safety and minimise environmental impact;(ii)
Simplicity: simple architecture, enhanced manufacturability; and(iii)
Schedule optimisation and constructability.

4. Main Design Features
Nuclear Steam Supply System
The NUWARDTM reactor is a fully integrated PWR reactor, housing asingle unique vessel containing all the main reactor coolant systemcomponents, including the CSGs, the pressuriser and the CRDM.

Reactor Core
The reference core is based on proven 17x17 (76) fuel assemblies usedin the operating PWR fleet with a shortened height and UO2 rods
(enrichment < 5wt% 235U). Due to the boron-free design, various 235U
enrichments and burnable poisons are used. The refuelling interval is
24 months for half the core.

Reactivity Control
The reactivity is controlled by means of control rods and solid
burnable poison. The reactor boron-free reactivity control allows for
simplification of auxiliary systems design and operation in bothnormal conditions and DBCs, as well as a drastic reduction in
effluents produced from operation.

Reactor Cross-section View

Reactor Pressure Vessel and Internals
The RPV and equipment layout are designed to facilitate enhanced in-factory manufacturing. A specific design
effort has been made to reduce the number of pipes connected to the RPV with the objective to limit the
maximum Loss Of Coolant Accident (LOCA) size to a 30 mm diameter break.

Reactor Coolant System and Steam Generator
The NUWARDTM reactor coolant system adopts an innovative steam generator technology using the plate heatexchanger concept. The CSGs are in direct connection to the reactor thus eliminate the need of external primaryloops. This makes the design highly efficient with a high thermal power per volume ratio. The overall size ofthe reactor coolant system is therefore significantly reduced given the reactor thermal power.

Pressuriser
The NUWARDTM pressuriser is integrated within the RPV head. The large volume of the pressuriser provides
margins for the operational transients as well as for normal operation of the reactor.

Primary pumps
Six (6) canned-rotor pumps are horizontally mounted onto the RPV, positioned under the CSGs in the cold leg
for efficient hydraulic conditions.

5. Safety Features
Engineered Safety System Approach and Configuration
The NUWARDTM reactor and associated safety systems are designed for: (i) Passive management of DBC
scenarios with no need for an external ultimate heat sink or any external electrical power supply (normal andemergency) for more than 3 days; (ii) Active management of Design Extension Condition (DEC)-A accidents,
with simple diagnosis and implementation of diversified systems; and (iii) In-Vessel Retention of the corium(IVR) strategy for DEC-B management.

44

The safety approach for NUWARDTM benefits from the following inherent features of the design to satisfy andmaintain a safe state that requires minimum intervention from the operating team: (i) Large reactor coolant
inventory (kg/MWth) providing inertia versus power transients; (ii) Integrated reactor coolant systemarchitecture reducing the maximum LOCA break size thus providing more coping time in the event of thedesign basis LOCA; (iii) Internal CRDMs preventing rod-ejection accidents; (iv) Boron-free in normal
operation (including all DBCs) providing large and constant moderator counter-reaction and preventing boron
dilution; (v) A metallic submerged containment providing passive cooling for several days, and (vi) A relatively
small core in a large vessel enabling the efficient implementation of the IVR concept for DEC-B accident
scenario.

Safety Approach and Configuration to Manage DBC
NUWARDTM incorporates 2 trains of passive heat removal, via a natural circulation transfer system of the
decay heat from the core to the water surrounding the unit containment (the reactor pool), through two
dedicated Safety CSGs (S-CSG) independent from the operational or the operational six CSGs. Each train canbe actuated by 2 diversified channels (diversified sensors and I&C) and redundant (backup spare) actuators.
The water surrounding the unit containment ensures the heat removal function for more than 3 days without
the need for an external ultimate heat sink. The passive vessel heat removal system is considered a D-passive
system according to IAEA classification. A set of 2 redundant low-pressure safety injection accumulators
provide the make-up of reactor coolant water inventory in case of LOCA. NUWARDTM includes safety featuresto prevent criticality risks. The use of an internal CRDM eliminates the occurrence of a rod ejection accident.
Dedicated safety systems to manage DBC are provided for each of the two reactor units.

Safety Approach and Configuration to Manage DEC
DEC-A systems include Low flowrate depressurisation system and active water injection. This systemprovides for the removal of the decay heat in case of a postulated common mode failure of redundant trains ofpassive DBC safety systems; high pressure borated water injection is available to cope with Anticipated
Transients Without Scram (ATWS) accidents. DEC-B systems include the low flowrate depressurisation
system; Flooding of the vessel pit in order to provide IVR of corium; and nitrogen injection to manage therisk of hydrogen combustion.

Containment System
The NUWARDTM design includes steel containment as the 3rd barrier, which is immersed in water. The
minimised LOCA break size and the efficient passive heat removal system result in a limited peak pressure
inside the steel containment which is passively cooled by the surrounding water. The containment is protectedagainst hydrogen accumulation risk in DBCs by passive recombiners.

Spent Fuel Cooling Safety Approach / System
Located between the 2 units that form the NUWARD™ design, is a shared spent fuel pool. The fuel assemblies
are moved through a transfer chute located at the top of the steel containment; this feature is available on each
unit. Along with the shared spent fuel pool is a shared refuelling machine used to achieve the fuel transfer.

6. Plant Safety and Operational Performances
The design target value for the lifetime capacity factor is above 90%, with the major planned refuelling only
outages scheduled for 20 days every 24 months. The reference refuelling strategy is to replace half of a core
every 2 years. The plant provides a storage capacity of spent fuel assemblies for 10 years of operation.

7. Instrumentation and Control Systems
The Instrumentation and Control (I&C) system designed for NUWARDTM is based on the defence in depth
concept, compliance with the single failure criterion and diversity.

8. Plant Layout Arrangement
The Nuclear Island (NI) building is located below ground (semi-buried) for protection against external hazards
and certain malicious acts; and increased ease of construction. The NI houses the 2 independent units andshared fuel storage pool. NUWARDTM plants are suitable for sea/lake-onshore and/or river-side sites, with
open-loop conventional condenser cooling, as well as inland sites with aero condensers. The basic gridinterface will be compliant with ENTSO-E and EUR requirements (typically 225kV/400kV and 50Hz).

45

NUWARDTM example plant layout

9. Testing Conducted for Design Verification and Validation
Various tests and studies are in progress to validate aspects of the design and its function, particularly aroundthe residual heat removal passive system. Design feedback and evolution as a result of the outcomes from thesestudies and all other aspects of design and verification will follow. Certain aspects of the design are drawn
from operational experience and design experience related to existing medium and high power PWRs.

10. Design and Licensing Status
NUWARDTM is close to completing its conceptual design phase at the time of writing and is preparing for prelicensing. A safety options file (Dossier d’Options de Sûreté or DOS) will be completed by the end of theconceptual design to be formally reviewed by the French safety authority (ASN). Following EDF’s initiative,
NUWARD™ is the case study for an on-going Joint Early Review, led by ASN with the participation of STUKand SUJB, respectively the Finnish and the Czech safety authorities, from which early insights into the design
development and safety approach are anticipated. Site permit for First Of A Kind (FOAK) is being pursued.
Agreement has been reached with the Government of France that a FOAK NUWARD™ will be built in France.
a number of potential sites are being considered for this.

11. Fuel Cycle Approach
The reference plant refuelling cycle is for half a core every 2 years. The plant provides a storage of spent fuelassembly for 10 years after operation before decommissioning.

12. Waste Management and Disposal Plan
Options regarding waste disposal are currently under assessment, taking note of industry best practice.

13. Development Milestones
2012 – 20162017 – 20192019 – 2022
2023 – 2026
From 2025
2026 – 2030
By 2030

46

Preliminary studies and technological innovation (using previously
developed patents).
Pre-conceptual design phase and technology validation
Conceptual Design Phase (and preparation for pre-licensing)
Basic Design Phase
Commercialisation
Detailed Design Phase
Target first concrete for the FOAK in FranceIMR (Mitsubishi Heavy Industries, Japan)
All content included in this section has been provided by, and is reproduced with permission of Mitsubishi Heavy Industries, Japan.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

Mitsubishi Heavy Industries,
Ltd. (MHI), Japan
Integral PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

1 000 / 350

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

15.5 / 5.0

Number of fuel assemblies
in the core
Fuel enrichment (%)

97

Core Discharge Burnup (GWd/ton)> 40

Refuelling Cycle (months)

26

Reactivity control mechanism

Control rods drive mechanism

Approach to safety systems

Hybrid (Passive + Active) system

Design life (years)

60

Plant footprint (m2)4900

RPV height/diameter (m)17 / 6

RPV weight (metric ton)–

Seismic Design (SSE)

0.3g

Fuel cycle requirements / Approach

Similar to existing PWR plants

Distinguishing features

Integral PWR with natural
circulation; employs two types ofin-vessel steam generator
Conceptual design completed

Design status

Natural circulation

329 / 345
UO2 pellet / 21x21 square

4.8

1. Introduction
The Integrated Modular Water Reactor (IMR) is a medium sized power reactor with a reference thermal output
of 1000 MW(t) producing an electrical output of 350 MW(e). The IMR is developed for potential deployment
after 2025. IMR employs the hybrid heat transport system (HHTS), a natural circulation system for the primaryheat transport. The in-vessel control rod drive mechanism (CRDM) is the primary means of reactivity control.
These design features allow the elimination of the emergency core cooling system (ECCS).

2. Target Application
The IMR is primarily designed as a land-based modular power station to generate electricity. Because of its
modular characteristics, it is suitable for large-scale power stations consisting of several modules and also forsmall distributed-power stations, especially for small grids. IMR can also be used for cogeneration of electricity
and district heating, seawater desalination, process steam production and so forth. IMR adopts structures,
systems and components that require no large-scale infrastructure. This facilitates regulatory authority’s
approval for the construction and operation of power plant.

3. Design Philosophy
The IMR is an integral PWR where the primary system components are installed within the reactor pressure
vessel (RPV). Main coolant piping and primary coolant pumps are eliminated by adopting natural circulation
system. Pressurizer is eliminated by adopting the self-pressurization system. There are two types of steam
generator (SG) inside the RPV; The first type is located in the vapour/upper region of the RPV and the other47

is located in the liquid/lower region of the RPV. These SGs are also used as decay heat removal heat exchangers
during normal startup and shutdown operations and in accidents. Hence to eliminate the need of ECCS, theSGs serve as a passive safety system that do not require any external power. The IMR has a moderation ratio
similar to the operating PWRs. Thus, its properties of fresh and spent fuel are similar. This allows forconventional safeguards measures and PWR management practices for new and spent fuel. Support systems,
such as the component cooling water system, the essential service water system and the emergency AC powersystem, are designed as non-safety grade systems, made possible by use of a stand-alone diesel generator.

4. Main Design Features
Nuclear Steam Supply System
The HHTS is employed to transport the fission energy released in the fuel to the SGs by both vapour formationand liquid temperature rise. The energy transported by vapour produces secondary steam in SGV, and theenergy transported by liquid temperature rise produces secondary steam in SGL. The SGV also has a functionof primary system pressure control, and the SGL has the function of core power control through the core inlet
temperature by controlling the feedwater flow rate.

Reactor CoreThe IMR core consists of 97 fuel assemblies in 21×21 array with an average enrichment of 4.95 % and produces
an output of 1000 MW(t). The refuelling interval is 26 effective full-power months. The power density is about
40% of current PWRs but the fuel lifetime is 6.5 years longer, so that an average discharged burnup of 46
GWd/ton can be attained, which is approximately the same as in current PWRs. The cladding material is Zr–
Nb alloy to assure integrity at a temperature of 345°C and over the long reactor lifetime. To maintain the core
thermal margin and to achieve a long fuel cycle, the core power density is reduced to one-third of thatconventional PWRs. The design-refuelling interval is three (3) years in three (3) batches of fuel replacement.
The fuel rod design is the same as that for a conventional PWR.

Reactivity Control
The chemical shim reactivity control is not used in the IMR, rather both control rods that contain enriched 10B
and burnable absorbers control the whole reactivity. Control rods with 90 wt% enriched B4C neutron absorber
perform the reactivity control, and a soluble acid boron system is used for the backup reactor shutdown to
avoid corrosion of structural materials by boric acid. The hydrogen to uranium ratio (H:U) is set to five, whichis larger than in conventional PWRs, to reduce the pressure drop in the primary circuit. The coolant boils in
the upper part of the core and the core outlet void fraction is less than 20% locally and less than 40% in thecore to keep bubbly flow conditions. To reduce axial power peaking caused by coolant boiling, the fuel consistsof two parts: the upper part with higher enrichment and the lower part with lower enrichment. Additionally,hollow annular pellets are used in the upper part of the fuel to reduce axial differences in burnup rate. Twotypes of in-vessel CRDMs are adopted. One is motor driven CRDM for the control bank. This CRDM has thefunction of controlling reactivity during operation by finely stepping the control rod position. The other is thehydraulic type CRDM. This CRDM has the scram function and applies to the shut-down bank. The control
rods connected to this CRDM are moved by hydraulic force from the bottom position to the top, and then heldby electro-magnetic force. When the scram signal is initiated, the control rods are released and inserted into
the core by gravity by turning off the power to the CRDM.

Reactor Pressure Vessel and Internals
The upper part of the RPV inside diameter is about 6 m in order to accommodate the in-vessel SGs. The inside
diameter of the lower part of the RPV is reduced to about 4 m in order to minimize the cold-side water volume.
In order to eliminate the necessity for the consideration of LOCA, the largest diameter nozzle connected to the
RPV is reduced to less than 10 mm. In addition, the lowest location of the nozzle is above the core to improve
the reliability of the RPV. The core is located in the bottom of the RPV and the SGs are located in the upper
part of the RPV. Control rod guide assemblies are located above the core and a riser is set above the control
rod guide assemblies to enhance the natural circulation.

Reactor Coolant System
In the HHTS, IMR employs natural circulation and a self-pressurized primary coolant system, altogether
resulting in a simple primary system design without reactor coolant pumps and pressurizer, it also reduces
maintenance requirements. The HHTS reduces the size of RPV. The coolant starts boiling in the upper part ofthe core, and two-phase coolant in bubbly flow flows up in the riser and condenses in the SGs. This design
approach increases coolant flow rate and thus reduces the height of RPV to transport the heat from the core.
The IMR primary cooling system design under bubbly flow makes it easy to employ PWR design technologies.

Steam Generator
The IMR adopts two types of SG. The first one is the SG in vapour region (SGV) located above the water levelin the RPV. The energy transported by vapour formation generates secondary steam through SGV. As thevapour in the RPV is condensed by SGV, controlling the feedwater flow rate to SGV controls the RPV
pressure. The other is the steam generator in liquid region (SGL) of the RPV. The energy transported by liquid48

temperature rise generates secondary steam through SGL. Because the core inlet temperature can be controlled
by the amount of heat removal through SGL, the core power can be controlled by feedwater flow rate to SGL.
By this method, the movement of the control rods for controlling reactor power will be minimized. For SGL,a U-type tube bundle is adopted, since it is necessary to minimize pressure drops on both the primary andsecondary sides to maintain good natural circulation performance. A C-type steam generator is adopted for
SGV to optimize space utilization in the vapour part of the RPV.

Pressurizer
The physical pressurizer vessel is eliminated by adopting the self-pressurization system.

5. Safety Features

Engineered Safety System Approach and Configuration
By adopting an integral type primary system, accidents that may cause fuel failure, such as loss of coolant
accidents (LOCA), rod ejection (R/E), loss-of-flow (LOF) and locked rotor (L/R), are eliminated in IMR. Since
the diameter of the pipes connected to the RPV is limited to less than 10 mm, the water level in the RPV canbe maintained at normal levels by water injection from the charging pumps. There are two trains of the SDHS.
If a malfunction such as SG tube leakage occurs, system functions are maintained. The capacity of chemical
and volume control system (CVCS) is provided via eight 3-inch pipes connected to the RPV. No ECCS andcontainment cooling/spray systems are required in IMR. Safety injection systems are eliminated by adopting
the SDHS and by limiting the nozzle diameter connected to the primary system. Hence, containment spraysystem is also eliminated. The auxiliary feedwater system is used for startup and shutdown procedures during
normal operation. The auxiliary feedwater system is not a safety system. When the auxiliary feedwater system
becomes unavailable, the SDHS is actuated. The IMR adopts simplified support systems, such as thecomponent cooling water system (CCWS), the essential service water system (ESWS) and the emergency AC
power system. These are designed as non-safety grade systems powered by a stand-alone diesel generator.

Decay Heat Removal System
The SDHS is activated to remove the decay heat from the RPV to the atmosphere. Even if water leakage occurs
and the charging pumps fails to operate, water leakage would be terminated automatically when the pressuresinside and outside the RPV are equalized. In the passive steam generator cooler (PSGC), decay heat is removed
by water-cooling in the early stage of the accident and then, the heat transfer mode is gradually replaced byair-cooling. Therefore, water, power and operators are not necessary for maintaining the plant safety.

Containment System
A compact containment vessel (CV) is made possible due to the integrated primary system and simplified
auxiliary systems. The IMR uses reinforced concrete containment. A higher design pressure of the containment49

is to meet the safety requirement that water leakage from RPV shall be terminated automatically. Since thisCV is about one size larger than the RPV, it is expected to resist high pressure. The reactor containment facility
is part of the engineered safety systems, which include SDHS. The containment system is designed to suppressor prevent the possible dispersion of large quantities of radioactive materials.

6. Plant Safety and Operational Performances
The IMR is designed to operate automatically within the range of 20 to 100 % of rated output power by thereactor control system. Even in the low output range below 20%, the control system can control the reactor
automatically in the low power-operating mode. The primary system pressure and reactor power are controlled
by feedwater and control rods.

7. Instrumentation and Control Systems
The instrumentation and control (I&C) systems provide the capability to control and regulate the plant systems
manually and automatically during normal plant operation and provide reactor protection against unsafe plant
operation. Fully digitalized I&C system including computerized control board for plant operator are provided
with the required conventional system.

8. Plant Layout Arrangement
The IMR concepts of building layout are reducing the bill of quantity for construction material, shortening theconstruction period and standardizing the plant design. Utilizing the steel plate reinforced concrete andsimplify the shape of building and structures achieve the construction cost and period reduction.

Reactor Building
Ground level is assumed to be flat land above sea level. The bedrock is assumed to be less than 40 m below
ground to enable the use of pile foundations. The integrated reactor building can house two units. Exclusionof waste disposal facilities in another building.

Balance of Plant
The advanced BOP system allows the utilization of produced heat for non-electrical applications such asprocess heat, mining (oil sand extraction) and desalination. The turbine generator, turbine, condenser, moistureseparator and reheater (MSR) and their auxiliary equipment are installed in the turbine building. The turbine
generator is arranged with its axis in line with the reactor.

9. Design and Licensing Status
The IMR conceptual design study was initiated in 1999 by MHI. A group led by MHI including Kyoto
University, the Central Research Institute of the Electric Power Industry and the Japan Atomic Power
Company developed related key technologies through two projects, funded by the Japanese Ministry of
Economy, Trade and Industry (2001–2004 and 2005–2007). Validation testing, research and development forcomponents and design methods, and basic design development are required before licensing.

10. Fuel Cycle Approach
The IMR fuel cycle approach including spent fuel management is in line with the approach for the existing
PWR plants. It leads the minor design modification for existing fuel cycle facilities, and the IMR approach isaccepted by public without discomfort.

11. Waste Management and Disposal Plan
The IMR waste management and disposal plan is in line with the existing PWR plants concept. The IMR
approach is accepted by public without discomfort.

12. Development Milestones
1999
2001-2004

2005-20072009-20112019

50

MHI started conceptual design study for IMR.
An industry-university group led by MHI, including Kyoto University, Central ResearchInstitute of Electric Power Industries (CRIEPI), the Japan Atomic Power Company(JAPC), and MHI were developing related key technologies through two projects, fundedby the Japan Ministry of Economy, Trade and Industry. In the first project, the feasibilityof the HHTS concept was tested through experiments.
In the second project, the thermal-hydraulic data under natural circulation conditions for
the HHTS design were obtained by four series of simulation tests using alternate fluids.
Startup transient tests to verify the startup ﬂow instability were studied
MHI is developing a new Small Reactor based on the IMR experiences with funding
support by the Ministry of Economy, Trade and Industry.

i-SMR (KHNP & KAERI, Republic ofKorea)All content included in this section has been provided by, and is reproduced with permission of, KHNP& KAERI, Republic of Korea.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
KHNP&KAERI
country of origin
Republic of Korea
Reactor type
Integral PWR
Coolant/moderator
Light water / Light water
Thermal/electrical capacity,
540 / 170
MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15.0

Core inlet/outlet coolant
temperature (oC)295.5 / 320.9

Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)UO2 pellet / 17×17 square
69

Reactivity control
mechanismApproach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

Control rod, burnable absorber rods,
moderator temperature
Fully passive80NA
23 / 5
650NA
650
0.5g
Conventional LWR requirements applied

Distinguishing features

At least 72 hours for core cooling without
AC/DC power
Conceptual design

Design status

< 5.0
24
< 62.01. Introduction
In order to promote sustainable development and respond to climate change, the i-SMR is on development in
Republic of Korea with improved safety, economy and flexibility compared to existing nuclear power plants.
The i-SMR is an innovative small modular PWR producing 540MW thermal power. The i-SMR has an integral
Reactor Coolant System (RCS) configuration that eliminates large pipes for connection of the major
components. The major primary components such as a core with 69 fuel assemblies, eight Reactor Coolant
Pumps (RCPs), a helical once-through Steam Generator (SG), and a Pressurizer (PZR) are installed in a single
Reactor pressure Vessel (RV) and equipped in a compact steel Containment Vessel (CV). Due to the integral
arrangement, the possibility of a Large Break Loss of Coolant Accident (LB-LOCA) is inherently eliminated.

2. Target Application
The i-SMR is a multi-purpose reactor for electricity production, heat supply for industries, and hydrogenproduction. The i-SMR is four module design, and its total electricity is comparable to substitute theconventional fossil power plant.

3. Design Philosophy
The i-SMR implements design simplification by integrating the primary coolant system and passive safetysystems without external safety water injection, least operating action by a high-level autonomous control.

51

4. Main Design Features
Nuclear Steam Supply System
The i-SMR is an integral type PWR with compact steel containment vessel. This configuration eliminates a
LB-LOCA from the design bases events. The Nuclear Steam Supply System (NSSS) consists of the reactor
core, SG, RCPs, Control Element Drive Mechanisms (CEDMs), and reactor internals in the RV. The primarycooling system is based on forced circulation by the RCPs during normal operation. The system has natural
circulation capability in emergency conditions that ensures 72 hours coping time for core cooling without
AC/DC power. Moreover, air cooling for decay heat removal after water source depletion provides forunlimited long-term cooling capability.

Reactor Core
The core of i-SMR consists of 69 fuel assemblies. The active height of fuel assembly is 2.4 m and 17x17 square
fuel assembly containing gadolinia mixed in uranium dioxide with less than 5% enrichment. The refuelling
scheme is a two-batch application with 24-month cycle length.

Reactivity Control
Reactivity control during normal operation is achieved by control rods and moderator temperature. The i-SMR
adopts boron-free operation and eliminates significant amount of component and maintenance activities related
to boron generation and concentration. Burnable poison rods are introduced to give flat radial and axial power
profiles, which results in an increased thermal margin of the core. The i-SMR adopts an internal control
element drive mechanism which excludes the rod ejection accident. A large number of control rods in i-SMR
core assures a relatively high control rod worth.

Reactor Pressure Vessel and Internals
The reactor vessel assembly of the i-SMR contains its major primary components such as a core, a SG, a PZR,
eight RCPs, and CEDMs in a single RV.

Reactor Coolant System
The RCS transfers heat generated from the core to the secondary system through the SG and plays a role of abarrier that prevents the release of reactor coolant and radioactive materials to the public. The RCS and its
supporting systems are designed with sufficient core cooling margin for protecting the reactor core fromdamage during all normal operation and Anticipated Operational Occurrences (AOO).

Steam Generator
The SG of i-SMR, which is installed inside the RV, is a helical once-through type. Feedwater flow upward
through the tube side of SG. The tubes of the SG are helical-designed to provide the maximum heat transfer
area in a limited space. The primary and coolant flow inside the RV is treated as single path for the normal
operation.

Pressurizer
The pressurizer for i-SMR is located at top-side of the RV, since this configuration have advantages forpressurizing the reactor coolant in a subcooled state. The pressurizer is designed as the steam pressurizer whichsaturated steam and water are co-existed in pressurizer region. The steam pressurizer has an advantage that thesimple control schemes are provided by two-phase phenomena during the transients.

Primary pumps
The i-SMR has eight RCPs vertically installed at upper part of the Reactor Closure Header (RCH). Each RCP
is an integral unit consisting of a canned asynchronous three phase motor. Rotational speed of the pump rotor
is measured by sensors installed in the upper part of the motor. Since a canned motor pump does not require
pump seals, possibility of the Small Break Loss of Coolant Accident (SB-LOCA) associated with a pump seal
failure is basically eliminated.

5. Safety Features
Engineered Safety System Approach and Configuration
Safety approach for design and operation of the i-SMR is based on the defense-in-depth philosophy. Multiple
barriers such as fuel pellet, cladding, RV, and CV prevent radioactive release to the environment and thosebarriers are protected by fully passive safety systems. The safety systems of i-SMR, a sensible mixture ofproven technologies and advanced design features, are designed to function automatically on demand without
critical operator action. Under a postulated design basis accident, the safety systems cool the RCS below thesafe shutdown condition within 36 hours and keep the core undamaged for 72 hours without any corrective
actions by operators. Heat from the RCS can be dissipated to the heat sinks, such as the Emergency Cooling52

Tank (ECT) and the Reactor Building (RB) atmosphere. Also the safety systems maintain the safe shutdown
condition for at least 72 hours under the complete loss of onsite and offsite AC power.

Emergency Core Cooling System
The Emergency Core Cooling System (ECCS) is composed of Emergency Depressurization Valves (EDVs)
and Emergency Recirculation Valves (ERVs). Three EDVs and two ERVs open when pressure difference
between the RV and CV reaches the predetermined setpoint. When LOCA occurs, the coolant releases to the
CV annulus between the RV and the CV through EDVs. When ERVs open, water flows into the RCS throughERVs due to the hydrostatic head difference and forms natural recirculation path.
Decay heat is removed by a Passive Auxiliary Feedwater System (PAFS) and Passive Containment Cooling
System (PCCS) in non-LOCA and LOCA scenarios, respectively.

Schematic diagram of the ECCS

PAFS and PCCS schematics

Decay Heat Removal System
After reactor trip, when normal decay heat removal mechanism utilizing the secondary system is not working
properly, the PAFS brings the RCS to safe shutdown condition within 36 hours after accident initiation andmaintains the safe shutdown condition for at least another 36 hours. The PAFS consists of inlet pipes, PAFS
tubes, outlet pipes, and actuation valves. In normal operation, actuation valves are closed and main steam flows
to the turbine. When the PAFS signal occurs, the PAFS actuation valve and main steam isolation valve areclosed. Hence, the steam generated in the SG flows from the main steam line to the PAFS tube and condensed
due to the cold ECT water and return to the feedwater line. Therefore, safety function is performed for at least
72 hours without any corrective action by operator or the aid of AC/DC power for the design basis accidents.
The safety function of the PAFS is maintained continuously for a long-term period, coupled with the ECCS
without refilling the ECT.

Containment Cooling System
The CV is composed of stainless steel, and divided to upper and lower regions of distinct diameters. The CV
is supported by support lugs located beneath the CV. Outer surface of the CV is exposed to the RB inner
atmosphere. The CV provides penetration of PCCS piping connected between the PCCS Heat exchangers (HX)
and the ECT, and main steam and feedwater piping, etc. Double isolation valves mounted on the CV are
installed to prevent LOCA outside containment events for the RCPB lines penetrating the CV.
The PCCS consists of inlet pipes, the PCCS HX which is located inside the containment, and return pipes.
When LOCA occurs, the hot steam contacts the PCCS HX outer surface and condensation occurs. Heatedwater inside the PCCS HX flow upward to the ECT and natural circulation flow path formed. In the PCCS
system, there are no actuation signal and operator action required to perform its safety function.

6. Plant Safety and Operational Performances
The safety of the i-SMR is achieved in a core damage frequency less than 10-9 and a large release frequency
less than 10-10 per year. The operating experience of KHNP have been applied to the i-SMR with the expectedavailable factor no less than 95%. Load following operation of i-SMR is simpler than that of conventional
PWR because boron-free operation of i-SMR with a large reactivity feedback effect minimizes the movementof control banks and boron concentration control is also not necessary. The daily load following performance
simulation of the i-SMR core shows that radial peaking factor, 3D peaking factor and the axial offset were53

satisfied within the design limit.

7. Instrumentation and Control System
The Human-Machine Interface System (HMIS) features an advanced control room and the digital control
systems. HMIS combined with human factor engineering for ensuring safety by minimizing the likelihood of
human error provides the capability to protect, control, and monitor the nuclear power plants. The Integrated
Main Control Room (IMCR) is provided with Human-Machine Interface (HMI) devices from which operator
action can be taken simultaneously to operate for four Integrated Power Cells safely under all operating
conditions (except MCR fire) and maintain it in a safe condition under all operating conditions.
The HMIS is developing to satisfy all regulatory requirements such as independence, redundancy, defense-indepth and diversity and to improve the economics and operability.

8. Plant Layout Arrangement
The i-SMR is designed to accommodate four (4) integrated reactors and containment vessels, structures,
systems and components. The power block of the i-SMR consists of the reactor building, the control building,the turbine generator building and the compound building.

Plant layout and arrangement

9. Design and Licensing Status
The i-SMR design is developed in conformity with Korean law, codes and standards for nuclear power plants
and safety principles

10. Fuel Cycle Approach
The i-SMR adopts the same open fuel cycle as operating PWRs. The fuel cycle of i-SMR is 24 months.

11. Waste Management and Disposal Plan
The i-SMR has several design solutions to minimize radioactive waste generation. All radioactive waste will
be processed with the conventional PWR waste treatment.

12. Development Milestones
2023
2028

54

Basic design completedStandard design approval

SMART (KAERI, Republic of Korea and
K.A.CARE, Kingdom of Saudi Arabia)
All content included in this section has been provided by, and is reproduced with permission of, KAERI, Republic of Korea and K.A.CARE,
Kingdom of Saudi Arabia.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
KAERI, Republic of Korea andcountry of origin
K.A.CARE, Kingdom of Saudi Arabia
Reactor type
Coolant/moderator
Thermal/electrical
capacity, MW(t)/MW(e)Integral PWR
Light water / light water
365 / 107

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15 / 5.8

Core inlet/outlet coolant
temperature (oC)

296 / 322

Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)UO2 pellet / 17x17 square
57

Reactivity control
mechanismApproach to safetysystems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

Control rod drive mechanisms and solubleboron
Passive60
90 00018.5 / 6.5
1070 (including coolant)> 0.3g with 0.18g of automatic shutdown
Conventional LWR requirements applied(spent fuel storage capacity: 30 years)
Coupling with desalination and process
heat application, integrated primarysystemDistinguishing features

Design status

<5
30
< 54

Detailed design

1. Introduction
The System-integrated modular advanced reactor (SMART) is an integral PWR with a rated electrical powerof 107 MW(e) from 365 MW(t). SMART adopts advanced design features to enhance safety, reliability andeconomics. The advanced design features and technologies were verified and validated during the standard
design approval review. To enhance safety and reliability, the design configuration incorporates inherent safety
features and passive safety systems. The design aim is to achieve improvement in economics through system
simplification, component modularization, reduction of construction time and high plant availability.

2. Target Application
The SMART is a multi-purpose reactor for electric power generation, desalination, district heating, and process
heat for industries. SMART has been developed to be suitable for small or isolated grids. SMART has a unit
output large enough to meet the demands of electricity and fresh water for a city population of 100 000.

3. Design Philosophy
The SMART design adopts an integrated primary system, modularization and advanced passive safety systems
to improve the safety, reliability and economics. Safety performance of SMART is assured by adopting passive
safety systems together with severe accident mitigation features. Improvement in economics is achieved
through system simplification, in-factory fabrication, reduction of construction time and high plant availability.

55

4. Main Design Features
Nuclear Steam Supply System
The SMART has an integral reactor coolant system (RCS) configuration that enables the elimination of a large
break loss of coolant accident (LBLOCA) from the design basis events. The nuclear steam supply system(NSSS) consists of the RCS forming a reactor coolant pressure boundary, secondary system, chemical andvolume control system (CVCS), component cooling water system (CCWS), passive residual heat removal
system (PRHRS), passive safety injection system (PSIS), automatic depressurization system (ADS),
containment pressure and radioactivity suppression system (CPRSS), etc.

Reactor Core
The low power density design with a slightly enriched UO2 fuelled core ensures a thermal margin of greaterthan 15%. In the core, there are 57 fuel assemblies of 2 m long, standard 17x17 square of UO2 ceramic fuelwith less than 5% enrichment, similar to standard PWR fuel. A two-batch refuelling scheme without
reprocessing provides a cycle of 870 effective full power days for operation.

Reactivity Control
Reactivity control during normal operation is achieved by control rods and soluble boron. Burnable poisonrods are introduced to give flat radial and axial power profiles, which results in an increased thermal marginof the core. SMART adopts a typical magnetic-jack type control rod drive mechanism which has been widelyused in the commercial nuclear power plants (NPPs).

Reactor Pressure Vessel and Internals
The RPV houses the reactor core, 8 steam generators (SGs), 4 canned motor reactor coolant pumps (RCPs),
25 control rod drive mechanisms (CRDMs) and reactor internals such as the core support barrel assembly andthe upper guide structure assembly.

Reactor Coolant System
The RCS transfers core heat to the secondary system through the SGs and plays a role of a barrier that prevents
the release of reactor coolant and radioactive materials to the reactor containment. The major components of
the RCS are a reactor vessel assembly containing the core, pressurizer space, SGs, RCPs, CRDMs, and related
pipes, valves, and instrumentations. The forced circulation flow of the reactor coolant is formed along the flow
path by the RCPs during normal operation. The RCS and its supporting systems are designed with sufficientcore cooling margin for protecting the reactor core from damage during all normal operation and anticipatedoperational occurrences (AOOs).

Steam Generator
The SMART has 8 modular type once-through SGs with helically coiled tubes to produce superheated steam
under normal operating conditions. The SGs are located at the circumferential periphery between the core
support barrel and RPV above the core to provide a driving force for natural circulation flow in emergency
conditions. The small inventory of the secondary side (tube side) water in each SG prohibits a return to power
following a main steam line break accident. In case of an accident, the SG can be used as the heat exchanger
for the PRHRS.

Pressurizer
The in-vessel pressurizer uses the free volume in the upper part of the RPV. The primary system pressureduring normal operation is maintained nearly constant due to the large pressurizer steam volume and the heater
control. Due to the large volume of the pressurizer, condensing spray is not required for load maneuvering
operation. The reactor over-pressure at the postulated design basis accidents can be reduced through theactuation of pressurizer safety valves.

Primary pumps
The RCPs are installed horizontally on the external wall of the reactor pressure vessel. It is a mixed flow pumpadopting a canned motor. It consists of the pressure retainer, the impeller and diffuser, the shaft assembly, andthe motor. No coupling is needed to connect the impeller shaft and the motor shaft. All of the pump parts are
enclosed by the pressure retainer. Therefore, there is no mechanical seal device to prevent the reactor coolant
from leaking through the pressure retainer. The component cooling water flowing in the helical tubes removes
the heat on the motor. The reactor coolant pump sucks the reactor coolant through the annulus between thecore support barrel and upper guide structure, and then it discharges the reactor coolant to the space above thesteam generator.

56

5. Safety Features
Engineered Safety System Approach and Configuration
Safety systems of SMART are designed to function automatically on demand. These consist of the PRHRS,
PSIS, and CPRSS. Additional safety systems include the ADS and pressurizer safety valves, and a severe
accident mitigation system.

Decay Heat Removal System
After the reactor is shutdown, when the normal decay heat removal mechanism utilizing the secondary system
is not operable for any reason, the PRHRS brings the RCS to a safe shutdown condition within 36 hours after
accident initiation and maintains the safe shutdown condition for at least another 36 hours. Therefore, thesafety function operates for 72 hours without any corrective action by operators for the postulated design basis
accidents. The safety function of PRHRS is maintained continuously for a long-term period when theemergency cooldown tank (ECT) is replenished periodically by a refilling system designed according toRegulatory Treatment of Non-Safety System (RTNSS) requirements.

Emergency Core Cooling System
The PSIS provides emergency core coolingfollowing postulated design basis accidents.
Emergency core cooling is performed using
4 core make-up tanks (CMTs) and 4 safety
injection tanks (SITs). Core coolinginventory is maintained through passive
safety injection from CMTs and SITs. The
CMTs that are full of borated water provide
makeup and borating functions to the RCS
during the early stage of an SBLOCA or nonLOCA. The top and bottom of CMT areconnected to the RCS through the pressurebalance line and safety injection line,
respectively. The safety injection function of
the PSIS is maintained long term as the SITs
are replenished periodically.

Containment System

Emergency Core Cooling System

The containment system is designed to contain radioactive fission products within the containment building
and to protect the environment against primary coolant leakage. This safety function is realized by the CPRSS
as a passive safety system. The containment system is composed of a lower containment area (LCA), an upper
containment area (UCA), an in-containment refueling water storage tank (IRWST), and the CPRSS, besides
these, it includes CPRSS Heat Removal System (CHRS). In case of main steam line break (MSLB) or LOCA,
some of the released energy is absorbed into the IRWST and the rest is removed to environment by the CHRS.
Fission products are scrubbed in the IRWST water. For combustible gas control, passive autocatalytichydrogen recombiners are equipped inside the LCA and UCA.

6. Plant Safety and Operational Performances
The core damage frequency of SMART is estimated at about 2E-7 per reactor year for internal events. Power
manoeuvring operation of SMART is simpler than that of large PWR because only a single bank movement
and small insertion is required to induce small reactivity change. This feature minimizes coolant temperaturechange, relatively high lead bank worth due to a small number of fuel assemblies and the short core heightleading to rapidly damping the xenon oscillation. The daily load following performance simulation of SMART
core shows that radial peaking factor, 3D peaking factor and the axial offset were satisfied within design limit.

7. Instrumentation and Control System
High reliability and performance of I&C systems is achieved using advanced features such as digital signal
processing, remote multiplexing, signal validation and fault diagnostics, and sensing signal sharing forprotection and control system. The ex-core neutron flux monitoring system consists of safety and start-up
channel detectors which are located within the RPV, and digital signal processing electronics. The in–core
instrumentation system consists of 29 detector assemblies which are developed as mini type for SMART withfour stacked rhodium self–powered neutron detectors.

57

8. Plant Layout Arrangement
SMART NPP has been designed to have aseawater intake structure and otherbuildings including chlorination building in
the yard. Power block accommodates
reactor containment and auxiliary buildings(RCAB), turbine generator buildings andone compound building shared by two unitsof SMART. The RCAB houses reactor
containment, auxiliary and fuel handlingareas to adapt the small and modular plant
concept. Reactor containment area consistsof the LCA and UCA. The LCA houses theRPV, CMTs, and SITs. Auxiliary area
houses emergency cooldown tanks, main
control room (MCR), electrical and control
facilities, and safety-related equipment
required to provide safe shutdown
capability. The balance of plant (BOP)
design consists of turbine generator
buildings and electric power systems.

INTAKE CHANNELINTAKE
STRUCTURE

TURBINEGERNERATOR
BLDG.

WATER
TREATMENTBLDG.

SWITCH YARDCONTROLBLDG.

INTAKE
STRUCTURE

STANDBY
DG BLDG.

UNIT2
REACTORCONTAINMENT
AND AUX.
BLDG.

TURBINEGERNERATORBLDG.

COMPOUND
BLDG.

UNIT1
REACTORCONTAINMENT
AND AUX.
BLDG.

STANDBY
DG BLDG.

ADMIN.
BLDG.
GUARD
HOUSE

Plant Layout Arrangement

9. Testing Conducted for Design Verification and Validation
The advanced design features of SMART were verified by a comprehensive technology validation programthat includes safety tests and performance tests. The safety tests consist of core critical heat flux tests, separate
effect and integral effect tests of the safety systems, thermal-hydraulic experiments, and digital man-machineinterface system (MMIS) tests. The performance tests cover fuel assembly out-of-pile tests, performance testsof the major components including RPV dynamic tests, RCP mockup test and SG irradiation test, and MMIS
control room tests.

10. Design and Licensing Status
Korea Atomic Energy Research Institute (KAERI) received the standard design approval for SMART fromthe Korean Nuclear Safety and Security Commission (NSSC) in July 2012. A safety enhancement program to
adopt passive safety system in SMART began in March 2012, and the testing and verification of the PRHRS
and PSIS were completed in the end of 2015. In September 2015, a pre-project engineering (PPE) agreement
was signed between the Republic of Korea and the Kingdom of Saudi Arabia for deployment of SMART. This
PPE project was successfully completed in February 2019. In 2019, Korea Hydro & Nuclear Power Co., Ltd.
(KHNP)/KAERI/K.A.CARE co-applied for standard design approval of the PPE design and First-of-a-Kind(FOAK) plant construction in Saudi Arabia will follow in due course.

11. Fuel Cycle Approach
The fuel cycle of SMART is 30 months. KEPCO-NF can provide SMART fuel with its fuel fabrication facility
increment schedule. The SMART spent fuels are stored in a spent fuel pool using storage racks. The current
storage capacity of spent fuel storage racks is 30 years which can be variable upon owner’s requirements.

12. Waste Management and Disposal Plan
The SMART has several design solutions to minimize radioactive waste generation. All liquid radioactive
waste will be processed through demineralizer package which can make the system design to be simple andminimize shipment of solid waste. Gaseous radwaste system provides sufficient holdup decay of radioactive
waste gases and release gases in a controlled manner. Solid radwaste system adopts polymer solidification
technology which can minimize shipped volume for spent resin.

13. Development Milestones
March 1999
March 2002
June 2007
July 2012
March 2012
September 2015
November 2015
February 2019
January 2020

58

Conceptual design development
Basic design development
SMART-PPS (Pre-Project Service)
Technology verification, Standard Design Approval (SDA)
First step of Post-Fukushima corrections and commercialization
Pre-project engineering agreement signed between Republic of Korea and Kingdom of
Saudi Arabia for the deployment of SMART in the Gulf country
Pre-Project Engineering started.
The Pre-Project Engineering completed.
SMART100 Standard Design Approval Applied.

RITM-200N (Afrikantov OKBM JSC,
Rosatom, Russian Federation)
All the content included in this section has been provided by and is reproduced with the permission of Afrikantov OKBM JSC, Russian Federation.

BASIC TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of origin
Afrikantov OKBM JSC, Rosatom,
Russian Federation
Reactor type
Integral PWR
Coolant/moderator
Water / Water
Thermal/electrical capacity,
190 / 55
MW(t)/MW(e)Primary coolant system circulation
Forced circulation
NSSS operating pressure
15.7 / 3.83 (steam pressure)(primary/secondary), MPa
Core inlet/outlet coolant temperature (°C)
283 / 321
Fuel type/assembly array
UO2 pellet/hexagonal assembly
Number of fuel assemblies in the core
199
Fuel enrichment (%)< 20
Refueling cycle (months)
60 – 72
Core discharge burnup (GW·d/ton)–Reactivity control mechanism
CRD mechanisms of CPS
Approach to safety system
Combined active and passive
safety systems
Design service life (years)
60
NPP footprint (m2)
~ 60 000, for NPP with 2 reactors~ 90 000, for NPP with 4 reactors
RPV height/diameter (m)7.5 / 3.4
RPV weight (metric ton)
164
Seismic design (SSE)
9 points on the MSK-64 scale
Requirements for or approach to theRefueling every 5 – 6 years
operating cycle
Distinguishing features
Modular design; integral reactor;
in-vessel corium retention insevere accidents
Design status
Detailed design

1. Introduction
The RITM-200 series reactors are the Afrikantov OKBM JSC’s latest development in the Generation III+ SMR
lineup. While possessing all the best characteristics of their predecessors, these reactors are based upon the timetested PWR technology and upon the Rosatom’s 400-reactor-year operating experience with small-sized reactors
on board icebreakers. Six RITM-200 reactors have been successfully installed on icebreakers Arktika, Sibir and
Ural. The lead icebreaker Arktika with two reactors on board was commissioned in October of 2020. The first in
the series, the icebreaker Sibir, was commissioned in January of 2022. For late 2022, it is planned to commence
operations of the icebreaker Ural. It is planned to construct 2 more icebreakers of this class – Yakutia andChukotka. The concept of and engineering solutions for the Afrikantov OKBM JSC-developed RITM-200
reactor plant are used in the RITM-200N reactor plant design for a pilot land-based small-sized nuclear power
plant.
The standardized design-schematic solutions and construction-layout solutions adopted in the RITM-200N
design and in the small-sized NPP design make it possible to implement a power lineup of energy-generating
sources – 55 MW (1 reactor plant + 1 STP in the main building); 110 MW (2 reactor plants + 2 STPs in the mainbuilding); 220 MW, 330 MW by a modular principle through constructing additional units with their own mainbuildings and cooling towers and with common plant-shared systems.

2. Target Application
The RITM-200N reactor plants may be used to generate electricity, cogenerate electricity and heat, and todesalinate sea water.

59

3. Design Philosophy
The reactor plant systems and equipment and nuclear fuel handling equipment were designed using the provenengineering solutions adopted by the nuclear propulsion plants in operation. The design is developed incompliance with the requirements of federal standards and regulations in nuclear energy use for NPPs. IAEA
requirements are incorporated. Enhanced safety requirements for Generation III+ reactors are met. Modularity isadopted for design and transport. An industrial large-unit installation process is used for construction.

4. Main Design Features
Nuclear Steam Supply System
The NSSS adopts an integral PWR. The reactor consists of a reactor core, four steam generators built in the
RPV, four canned-motor RCPs, control rod drive mechanisms of the CPS. The primary coolant system uses
forced circulation during normal operation and ensures natural circulation under accident conditions.

Reactor Core
Low-enrichment fuel assemblies are used as in KLT-40S. This ensures long-time operation without refuelingand meets the international non-proliferation requirements. The core height is 1650 mm. The core comprises anarray of 199 fuel assemblies.

Reactivity Control
Control rods are used to control reactivity. A group of CPS control rod drive mechanisms is provided tocompensate for the excessive reactivity during a startup, power operation and a reactor scram.

Reactor Coolant System
The primary coolant heat is transferred in steam generators sitting in the reactor pressure vessel. The steam
generators generate superheated steam from feed water and transfer the steam to the steam turbine plant (STP).
The secondary coolant system consists of four loops that include a steam generator, steam and feed water piping,valves, automatic safety devices and measuring instruments. The steam generators provide steam with atemperature of 295 °C, pressure of 3.83 MPa and an output of 305 t/h.

Steam Generator
The RITM-200N reactor plant design uses once-thorough straight-tube SG. The configuration of the steam
generating cassettes makes it possible to compactly place them in the reactor pressure vessel. The SGs are
divided into four loops of the primary coolant system. Each loop consists of three cassettes (making up the totalof 12) with shared feed water and steam headers. Each cassette contains 7 modules.

Pressurizer
The RITM-200N reactor plant uses an external gas pressurizer system that is well proven in the Russian marine
power engineering. The system is known for its design simplicity, which enhances reliability, ensurescompactness and does not require any electricity. The pressurizer system is divided into two independent groups
to reduce the pipeline diameter in the reactor nozzles and to reduce the coolant leakage in large break LOCAs inthe system piping. This solution allows one of the pressurizers to be used as a hydraulic accumulator, whichsignificantly enhances the reactor plant reliability in possible loss-of-coolant accidents.

5. Safety Features
The safety concept of the RITM-200N reactor plant is based upon the defense-in-depth principle in combination
with the inherent safety features and the use of active and passive systems. The inherent safety features areintended to limit the core power output as a function of the primary coolant pressure and temperature, the heatgeneration rate, the primary coolant system leak volume and outflow rate, the failed fuel fraction, the maintained
RPV integrity in severe accidents. The RITM-200N reactor plant optimally combines passive and active safetysystems, which cope with the events associated with abnormal operation, design-basis accidents and beyonddesign-basis accidents.

Approach to and Configuration of the Engineered Safety System
The high safety level of the RITM-series reactors is achieved through inherent safety features and through a
combination of passive and active safety systems. In addition to that, it is provided that equipment and channels
in the safety systems be redundant and functionally and/or physically separated to ensure high-level reliability.
The control rods of the CPS fall into the core by gravity or driven by a spring when the power supply is
disconnected from the electromagnetic couplings, which is followed by a reactor shutdown even in the case of acomplete blackout.

60

Residual Heat Removal System
The RHRS consists of four safety channels: an active safety loop with forced circulation through the steam
generator; an active safety loop with forced circulation through primary-to-third coolant system heat exchangerof the primary coolant purification system; two passive safety loops with natural coolant circulation throughsteam generators from water tanks. The water evaporated in the SGs condenses in the air-cooled heat exchangers
and comes back to the tanks with water-cooled heat exchangers. When the water is completely evaporated fromthe tanks, the air-cooled heat exchangers continue the cooling for an unlimited time. The combination of aircooled and water-cooled heat exchangers allows the overall dimensions of the heat exchangers and water tanks
to be minimized. All safety channels are connected to different SGs and ensure that residual heat be removed incompliance with the single failure criterion.

Emergency Core Cooling System (ECCS)
The ECCS consists of a safety injection system (SIS) to inject water into theprimary coolant system in order to mitigate the consequences of a LOCA
resulting from a pipeline break. The system is based on active and passive
safety principles with redundant active elements in each channel. The ECCS
consists of: Two pressurized passive hydraulic accumulators; Two active
channels with water tanks and two makeup pumps in either channel. In
combination with the residual heat removal system, the passive safety
channels provide a 72-hour grace period without any actions by personnel orupon a loss of power in a LOCA plus a complete blackout.

Containment System
The reactor plant is placed in a leak-tight enclosure in the form of a steel
containment. The containment includes three levels: (i) the first containment isshaped as a cylinder 8.7 m in diameter and 22 m tall. It is designed for internal
overpressure of up to 0.9 MPa, and is placed around the reactor pressure vessel
to isolate a possible leak of radioactive products; (ii) the second containment is
a strong containment of the building. It is made of thick reinforced concrete
walls (800 mm thick) to protect the first containment from external events;(iii) the third level of the containment is a building structure of reinforced
concrete walls to dissipate most of external impact energy and to minimize theimpact to the second containment. The strong containment design anddestructible elements take into account the maximum potential external events
including a crash of large commercial aircraft.

6. Plant Safety and Operational Characteristics
In the development of the RITM-200 plants and nuclear energy generating sources equipped with the RITM200-type plants, the priority area is preventing abnormal operation and accidents with account of the developing
and operating experience in marine plants and nuclear generating stations. The RITM-200N design is developedin conformity with Russian laws, standards and rules for nuclear power plants; in conformity with the safety
principles developed by the world community; and in conformity with IAEA recommendations.

7. Instrumentation and Control System
To monitor and control plant processes, an automated control system is provided in the small-sized NPP fittedwith the RITM-200N. In terms of safety functions, this system has the necessary redundancy and ensures bothautomated and remote control of the power plant.

8. General Layout of the Plant
The basic option of the small-sized NPP includes two RITM-200N reactors with a total installed electric powerof 110 MW. The basis for the placement of buildings and structures on the site is based on the principle ofzoning, which ensures maximum separation of buildings and structures according to their functional purpose incompliance with sanitary and fire gaps between buildings. Below is a general view of the small-sized NPP.
The reactor building and turbine building constructed on additional areas allow a gradual growth of energy
generation in 110 MW increments. In this case, the site area for a 100 MW small-sized NPP is 27 acres(0.11 km2); for 220 MW, 38 acres (0.15 km2); for 330 MW, 49 acres (0.19 km2).

61

9. Testing to Check and Validate the Design
The engineering solutions used in the reactor plant are traditional for marine power engineering. The solutions
have been tested in the course of many operating years and ensured the required reactor plant reliability andsafety performance. The RITM-200N reactor pertains to integral-type reactors. Integral-type reactors are used ina series of Project 22220 multipurpose nuclear-powered icebreakers Arktika, Sibir and Ural.

10. Design and Licensing Status
The small-sized NPP fitted with the RITM-200N reactor plant is at a design stage. A decision has been made toconstruct a pilot small-sized NPP in the Arctic zone near the town of Ust-Kuiga, Ust-Yansky ulus, the Republicof Sakha (Yakutia), Russia. The declaration of intent to invest into the construction of the small-sized NPP basedon the 55+ MW RITM-200N reactor plant in Ust-Yansky ulus, the Republic of Sakha (Yakutia), Russia wasapproved by Rosatom Director General A.Ye. Likhachev. It is planned to obtain the site license in March of2023.

11. Approach to the Operating Cycle
In the small-sized NPP, the nuclear fuel handling system uses a refueling complex analogous to that developed
for the multipurpose nuclear-powered icebreakers fitted with the RITM-200 reactor plant. The refueling complex
is used to load the core into the reactor and to unload the core to the spent fuel pool (SFP), to load spent fuelassemblies (SFAs) into shipping containers to transport them for reprocessing.

12. Waste Management System and Waste Disposal Plan
The design provides for the appropriate systems and equipment for radwaste handling (to convert liquidradwaste into a solid phase, to shred solid radwaste, to pack it in special containers) and temporary storage at thesmall-sized NPP.

13. Development Milestones
2018
2023
2024
2026
2027

62

A conceptual design of the land-based small-sized NPP
A site license for the small-sized NPP
A construction license; start of the basic SMR period
Operation license to be issued
First electricity to be generated by the small-sized NPP

VK-300 (NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of NIKIET, Russian Federation.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
NIKIET, Russian Federationof origin
Reactor type
Simplified passive BWR
Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

750 / 250

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

6.9

Number of fuel assemblies in
the core
Fuel enrichment (%)

313

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

41.4

Natural circulation

190 / 285
UO2 pellet/hexahedron

4

72

Reactivity control mechanism

Rod insertions

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

40 000RPV height/diameter (m)13.1 / 4.535

RPV weight (metric ton)

325

Seismic Design (SSE)

Max 8 point of MSK-64

Fuel cycle requirements /ApproachDistinguishing features

Once through fuel cycle with UO2Design status

Innovative passive BWR based on
operating prototype and welldeveloped equipment
Detailed design of reactor andcogeneration plant standard design

1. Introduction
The VK-300 is an integral simplified passive boiling water reactor (BWR) with a rated output of 750 MW(t)or 250 MW(e), adopting natural circulated primary coolant system. The design and operation of the VK-50
simplified BWR reactor in the Russian Federation for 50 years is used as a basis for the design of the VK-300
reactor. The design is based on a proven technology, utilizing the components developed and manufactured
for other reactor types. The VK-300 uses the reactor pressure vessel and fuel elements of the WWER-1000
reactor. The design configuration incorporates inherent and passive safety systems to enhance safety andreliability. The design aims to achieve improved economics through system simplification. The reactor core iscooled by natural circulation of coolant during normal operation and in emergency condition. The design
reduces the mass flow rate of coolant by initially extracting moisture from the flow and returning it to the core
inlet, ensuring a lower hydraulic resistance of the circuit and raising the natural circulation rate. The VK-300
reactor has a reactivity margin for nuclear fuel burnup due to the partial overloading and use of burnable
absorbers. The integral arrangement of reactor components and availability of preliminary and secondary
containments are non-proliferation features of VK-300.

2. Target Application
VK-300 reactor facility is specially oriented to the effective co-generation of electricity and heat for district
heating and for sea water desalination, having excellent characteristics of safety and economics.
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3. Design Philosophy
Design of the VK-300 is based on the proven WWER technologies and takes over the operating experience ofthe reactor of smaller size namely VK-50 that has successfully operated in Russian Federation over the last 50
years. Therefore, the enhanced reliability and economics are achieved by the use of some proven modified
structures and components in the design.

4. Main Design Features
Nuclear Steam Supply System
In a cogeneration plant with VK-300 reactor, steam goes directly from reactor to a turbine. After passing
several stages, some steam is extracted from the turbine and sent to the primary circuit of the district heat
supply or to the sea water desalination facility. Heat from the secondary circuit of the district heat facility issupplied to consumers. The circuit pressures are chosen so as to exclude possibility of radioactivity transport
to the consumer circuit.

Reactor Core
The hexahedron fuel assembly (FA) is formed by 107 UO2 ceramic fuel rods with enrichment of less than 4%
similar to VK-50 WWER fuel. There are 313 FAs in the core. Fuel burnup is 41.4 GWd/ton.

Reactivity Control
The reactor is provided with two independent reactivity control systems that use different principles of action.
The first system is a traditional rod system including 90 drives of the CPS. Each of the drives simultaneously
moves control rods installed in three adjoining fuel assemblies of the core. The second reactivity control system
is a liquid system intended for injection of boric acid solution to the reactor coolant at failures of the rod
reactivity control system. The system consists of pressurized hydraulic accumulators with a boric acid solution.
A lifting tube unit provides a guiding structure for the reactor control rods, which is very important at the upper
location of the CPS drives. The VK-300 reactor has a small reactivity margin for fuel burnup that creates preconditions for designing a simpler CPS system with light rods, which mitigates the consequences of accidents
with the CPS rod withdrawal.

Reactor Pressure Vessel and Internals
The VK-300 reactor vessel is a WWER-1000 reactor vessel in terms of external dimensions and material.
The VK-300 reactor includes the following internals:- a shell with the basket of the core;- a traction pipe unit;- a separator unit.
The traction pipe unit is an assembly of 90 vertical traction pipes of triangular-oval section and 25 circularpipes. The separator unit consists of 133 axial centrifugal separators.

Reactor Coolant System
The VK-300 primary cooling mechanism under normal operating condition and shutdown condition is bynatural circulation of coolant. The VK-300 design adopts an advanced coolant circulation system and amultistage separation in the reactor. A lifting tube (chimney) unit forms the raising and downstream coolant
flows, preliminary separates moisture and build-up the water inventory (between lifting tubes) that
immediately goes back to the reactor core in the event of the reactor shutdown or during accidents.

Steam Generator
The VK-300 reactor employs in-vessel cyclone separators that are designed and experimentally optimized to
be used in the vertical steam generators of the WWER-1000.

5. Safety Features
Innovative feature of the VK-300 project is the application of a metal lined primary containment (PC) of
reinforced concrete. The PC helps to provide safety assurance, economically and reliably using structurally
simple, passive safety systems.
The emergency cooldown tanks (ECTs) are located outside of the PC and are intended to function as
accumulators and primary inventory make-up. If there is a line rupture and the pressure of the PC and reactor
equalize, the ECTs actuate by gravity and fill the PC.
The residual heat is passively removed from the reactor by steam condensers located in the PC around thereactor that are normally flooded with the primary circuit water. When the level in the PC drops, the connecting
pipelines to the condensers are opened, the reactor steam condenses and returns back to the reactor. The
condensers are cooled with water from the ECTs.
At the same time the power unit design stipulates that the whole power unit will be within a leak-tight enclosure(the secondary containment). The containment accommodates the PC with the VK-300 reactor, emergency
cooldown tanks, turbine, spent fuel storage pools, refuelling machine and central hall crane. The containment
leak rate is 50% of the volume per day with the design pressure of not more than 0.15 MPa. Thanks to new64layout concepts for the main equipment of the VK-300 power unit, the containment dimensions do not exceed
the dimensions of the VVER-1000 reactor containment.

Engineered Safety System Approach and Configuration
The main technological solutions of VK-300:- single-loop reactor with natural coolant circulation;- power self-limitation due to negative reactivity and thermal coefficients;- passive removal of residual heat;- placement of reactor, turbine, emergency cooling tanks, spent fuel storage pool, reloading machine andcentral hall crane under a single secondary containment;- two independent power control and reactor shutdown systems (CPS using absorbing rods and CPS usingrods and boron fluid);- fully integrated reactor layout.

Decay Heat Removal System
The primary goal following a scram actuation is to remove residual heat from the shutdown reactor and ensure
its normal cool down. This function is performed by the residual heat removal system (RHRS) that passively
removes heat from the reactor in special heat condensers located inside the PC. The condensers are connectedto the reactor by pipelines that are filled with water during normal operation of the reactor. As the water level
decreases in the reactor, the upper pipeline opens for the steam passage from the reactor to the condensers andthe resultant condensate goes back to the reactor. The RHRS condensers are cooled with water from theemergency cool down tanks. The system is fully based on passive principles of action and ensures natural heattransport from the reactor to the emergency cool down tanks. The heat capacity of the tanks as such is enoughfor independent operation throughout the day (i.e. without personnel interference). This interval may be
prolonged for an infinite period of time due to the operation of the heat removal system from the tanks to theultimate heat sink. This is a simple and reliable system consisting of two heat exchangers connected withpipelines. One of the heat exchangers is plunged into the emergency cooldown tank water and the other isinstalled in the atmospheric air flow outside the reactor hall. The coolant in the system is water circulating in
the circuit naturally without pumps.

Emergency Core Cooling System
The emergency cooldown tanks contain the water inventory for emergency reactor flooding and core cooling
during steam or water line ruptures within the PCS. The emergency cooling tanks (ECTs) performs thefunctions of: (a) accumulating the reactor energy with the potential of transferring it to the end absorber for an
unlimited period of time; (b) compensating the cooling water inventory in the reactor during accidents byreturning the condensed coolant to the reactor; and (c) receiving steam or steam-water mixture (e.g., theexhaust of the reactor safety valves installed inside the PC). During a LOCA (rupture of a steam line or
feedwater pipeline adjoining the reactor within the containment), pressure increases inside the PCS which
serves as a signal for actuation of the reactor scram and passive closure of shutoff devices (valves) cutting thereactor off the external steam-water lines. A pressure reduction in the reactor as a result of coolant leak throughthe rupture creates conditions for the water delivery from the ECTs to the reactor via a special pipeline under
the action of hydrostatic pressure. The steam-air mixture goes via discharge pipelines from the containment to
the ECTs where it is condensed. As a result, a circulation circuit of the ECT – reactor –PCS – ECT is formed
and its function ensures long-term passive cooling of the reactor.

1-Feedwater
2-Out-core mixing chamber
3-Preliminary separation chamber
4-Pre-separated water outlet
5-Steam
6-Major separated water stream

Containment System
The VK-300 reactor adopts a metal-lined primary containment system (PCS) of reinforced concrete. The PCS
helps to solve the safety assurance problem economically and reliably using structurally simple passive safetysystems. The PCS is rather small, with volume about 2000 m3. The PCS of the VK-300 performs the functionsof: (a) a safeguard reactor vessel; (b) a protective safety barrier limiting the release of radioactive substances
during accidents with ruptures of steam, feedwater and other pipelines immediately near the reactor; and (c)providing the possibility of the emergency core cooling by the reactor cooling water making additional water
inventory unnecessary.

65

6. Plant Safety and Operational Performances
A set of reactor facility safety features and the concept of defense-in-depth against radioactivity escape allow
plant location in the vicinity of a residential district limiting the control area around the VK-300 cogeneration
plant by the dimensions of the cogeneration nuclear power plant (CNPP) site.

7. Instrumentation and Control Systems
Instrumentation and Control Systems basedon proven technologies, ensure cogeneration
NPP with effective operation and provide
safety assurance.

8. Plant Layout Arrangement
The turbine-generator system was developed
to produce 250 MW(e) electricity incondensing mode and heat of up to 465
MW(t) within a nuclear cogeneration plantfor district heating and for sea water
desalination. The VK-300 turbine is mainly
based on an element of the WWER-1000
turbine. Heat production systems were
designed to supply heat with no radioactivity.

Reactor and Turbine Building
Given the necessity of deploying cogeneration NPP within the city limits, with regard for the single-circuit
layout and the necessity of raising the reliability of the environmental protection during accidents, the power
unit design stipulates that all of the power unit will be within a leak-tight enclosure (the containment). The
containment accommodates the PC with the VK-300 reactor, emergency cooldown tanks, turbine, spent fuelstorage pools, refuelling machine and central hall crane. The electric generator is installed in a separate annex
outside the containment using a shaft that passes through the containment wall to beyond the containment. The
containment is an attended room whose primary function is to protect the reactor from external impacts suchas aircraft fall, terrorist acts, etc. Thanks to new layout concepts for the main equipment of the VK-300 power
unit, the containment dimensions do not exceed the dimensions of the WWER-1000 reactor.

Electric Power System
Electric Power System of VK-300 cogeneration power unit based on 220 MW(e) turbogenerator.

9. Design and Licensing Status
Research and development activities are currently under way for further validation and actualization of thedesign approach adopted in the VK-300 design.

10. Fuel Cycle Approach
The standard fuel cycle option for the VK-300 is a once-through fuel cycle with uranium dioxide fuel.
According to the design of the VK-300, spent fuel assemblies should be stored in the cooling pond for 3 years
after discharge from the reactor core and then transported to the fuel reprocessing plant without further longterm on-site storage. The standard fuel reprocessing method as used for WWER-1000 type reactors.

11. Waste Management and Disposal Plan
Radioactive waste is to be transferred to the National Radioactive Waste Management Operator for subsequentdisposal.

12. Development Milestones
1998
2002
2003
2004
2009
2013

66

Conceptual design development
Detailed design development
Cogeneration plant conceptual design development
Feasibility study of the pilot cogeneration plantFeasibility study of pilot cogeneration plant upgrade
Design validation, actualization and commercialization

KARAT-45 (NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of NIKIET, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

NIKIET,
Russian Federation
BWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

180 / 45-50

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

7.0 / Number of fuel assemblies in
the core
Fuel enrichment (%)

109

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

45.9

Reactivity control mechanism

Control rods drive

Approach to safety systems

Passive

Natural circulation

180 / 286
UO2 pellet/hexagonal

4.5

84

Design life (years)

80

Plant footprint (m2)9000

RPV height/diameter (m)11.15 / 3.10

RPV weight (metric ton)

176

Seismic Design (SSE)

0.3g

Fuel cycle requirements /Approach

Refueling (fuel shuffling) interval is
up to 800 EFPD; Fuel assembly life
cycle is about 6.6 years
Designed for extreme arctic andnorthern area conditions
Conceptual Design

Distinguishing features
Design status

1. Introduction
KARAT-45 is a small boiling water reactor (BWR) with a rated power of 45 MW(e) designed by NIKIET as
an independent cogeneration plant for producing electric power, steam and hot water. It is developed as thebase facility for the economic and social development of the Arctic region and remote extreme Northern areas67

of Russian Federation.

2. Target Application
KARAT-45 power unit has a high load follow capability to cope with daily power variation from 20% to 100%of nominal capacity.

3. Design Philosophy
The BWR technology was selected as a basis for the design and technology development of KARAT-45 dueto the following rationales: BWR employs single circuit removal of heat so capital cost for construction canbe minimized; lower system pressure poses fewer challenges to the reactor vessel; BWR has inherent selfprotection and self-control properties due to negative void and temperature reactivity coefficients. KARAT45 complies with Russian regulatory requirements and IAEA guidelines. The primary cooling mechanism forthe reactor core is natural circulation for all operating modes. The reactor vessel because of its small size will
be shop-fabricated in modular fashion to make it transportable. The reactor is designed for a long service life.

4. Main Design Features
Nuclear Steam Supply System
The reactor employs a single-circuit heat removal system. The reactor steam removal system is designed totransport the steam generated in the reactor to the turbine. The system removes heat from the reactor during
the reactor start-up, power operation and shutdown, as well as in some operational events when this systemand the feed water supply system are serviceable. In normal conditions of the reactor power operation,
saturated steam is fed from the reactor to the turbine via two steam lines. Isolation valves of the primary circuit
leak proof enclosure formed by the primary containment are installed immediately in front of and behind thecontainment penetration. The isolation valves are opened during normal operation. KARAT-45 reactor’s
design is based on gravity-type steam separation without centrifugal axial separators, which adds to the reactor
safety and makes it different from some other similar type reactors of bigger scale.

Reactor Core
The reactor core is located in the lower part of the reactor and consists of 109 fuel assemblies (FA). The corehas five complete FA rows and an incomplete sixth row. There are six steel reflector blocks at the core
periphery for the vessel protection against radiation. The FAs are installed inside the support grid locations. In
the upper part, the FAs are arranged in a hexagonal lattice with a pitch of 185 mm. Control orifices are installedat the core inlet for the coolant flow hydraulic profiling.

FA
Reflector
Core barrel
Reactor vessel

Reactivity Control
The core includes 109 Control and Protection System (CPS) rods fitted in the FA guide channels. Functionally,
the rods are divided into 4 emergency protection rods and 105 control rods. A rod is a shroud less structure
consisting of eight cylindrically shaped absorber elements. The absorber material is boron carbide (B4C). The
control rods are grouped into clusters to reduce the number of actuators. One actuator is used to move three,
two or one rod.

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) accommodates the reactor core and the reactor internals, including two
feedwater supply headers, two emergency heat exchangers and a block of louver-type separators. The reactor
vessel has an elliptical bottom, cylindrical shells, and two nozzle and main connector flange shells welded one
to the other. The RPV outer diameter is 3100 mm, wall thickness 100 mm, and height 11 150 mm.

Reactor Coolant System
The reactor has a single-circuit heat removal system. The coolant circulation is natural. The coolant (lightwater) flows through the reactor core upwardly while being heated and boiled. The steam generated after68

drying is fed to the turbine before being discharged into the condenser downstream of the turbine. Some partof the steam could be extracted from the turbine for heating the plant’s in-house water and for heat supply.
After leaving the condenser, the water is pumped through heaters and enters the deaerator. Using feedwater
pumps, the deaerated condensate is fed into the reactor through feedwater supply headers. Inside the reactor,
the feedwater is mixed with leftover water from steam separation and fed to the core inlet.

Steam Generator
Steam is generated directly inside the reactor vessel and, having been dried in the steam space, is fed into theturbine with a humidity of not more than 0.1 wt.% using gravity separation and louver-type separators built in
the reactor. Downstream of the turbine, the steam is dumped into the condenser.

Pressurizer
There is no pressurizer. Pressurization is achieved through negative feedbacks on temperature, power and void
reactivity effects.

5. Safety Features
The safety concept of the KARAT-45 reactor is based on inherent self-protection features, the defence-indepth approach and a system of barriers to the release of radioactive materials into the environment. The
concept is aimed at preventing accidents and mitigating their consequences, should these occur. To achievethis, normal operation systems and safety systems are required to perform reactivity control, core cooling andconfinement of radioactive materials in the required limits.

Engineered Safety System Approach and Configuration
One of the major principles of safety systems design is the requirement that they should operate at any designbasis initiating event and during failure of any active or passive component with mechanical parts
independently on the initiating event (single failure principle). The safety system design also meets therequirement for the systems to perform its functions automatically and reliably with the smallest possible
number of active elements involved and using the passive protection principle.

Decay Heat Removal System
The decay heat removal system is designed to remove heat from the reactor core during unexpected operational
occurrences and events caused by a loss of heat removal due to the feedwater supply and steam discharge
systems failure. The system ensures the nuclear fuel cooling function. The system is based on a passive
principle of action with heat removed from the reactor through natural circulation.

Emergency Core Cooling System
The emergency core cooling system is designed to supply the in-vessel natural circulation circuit with water
during accidents with loss of the primary circuit integrity. The system uses passive principle of action toorganize the coolant movement. The emergency mitigation of the primary coolant loss is ensured passively bydraining water from the emergency cooldown tanks into the reactor due to the gravitation because of thedifference in the tank and reactor elevations.

Containment System
KARAT-45 reactor is located inside a reinforced concrete containment with a stainless steel lining. The
containment serves to localize accidents and is designed to withstand a pressure of up to 3 MPa. It forms anadditional barrier to the leakage of radioactive materials into the environment while limiting, by its volume,the coolant loss during a reactor vessel break. There are isolation gate valves installed on pipelines at thecontainment outlet.

6. Plant Safety and Operational Performances
The major objective of the safety assurance arrangements is to limit the KARAT-45 radiation impacts on the
personnel, local population and the environment during normal operation, anticipated operational events andaccidents. KARAT-45 reactor features the following inherent self-protection properties:- Negative temperature and void reactivity coefficients;- Passive cooling of the reactor core based on natural coolant circulation both during normal operation andanticipated operational violations;- Sufficient amount of water in the emergency cooldown tank (ECT) for long-term decay heat removal;- Moderate thermal power density of fuel elements and reliable removal of residual core heat by merely fillingthe core with coolant;- A substantial amount of coolant above the reactor core to ensure the reliable fuel cooling in majority ofpossible emergencies;- Self-limitation of the in-vessel pressure variation rate due to the damping properties of the steam blanket.

7. Instrumentation and Control Systems
In-core monitoring system is designed for monitoring of thermal-hydraulic and neutron properties of the69

reactor core and in-core coolant natural circulation flow which are measured directly or indirectly in different
operating modes of the reactor.
The following parameters are expected to be monitored in the KARAT-45 reactor:- Continuous monitoring of neutron parameters defining the reactor period, neutron power and the control
rod position;- Continuous monitoring of thermal parameters defining the reactor’s thermal power, the reactor water level,
and temperature of water at FA inlet, in steam space and at the reactor vessel surface;- Periodic water chemistry control;- Periodic inspection of the thermal reliability of the reactor core operation. In-core power density field
monitoring.

8. Plant Layout Arrangement
The building layout plan for the land-based power unit of KARAT-45 is designed in such a way that the reactor
system, including its servicing systems, spent fuel pool, and auxiliaries are located in double protective aircrash resistance buildings. The overall weight and size parameters of the reactor unit due to its modular natureand transportability allow the delivery of unit assembled at factory directly to the construction site by railway
or other means of transportation.

9. Design and Licensing Status
KARAT-45 design was developed in conformity with Russian laws, norms and rules for land-based NPPs andsafety principles developed by the world community and IAEA recommendations.

10. Fuel Cycle Approach
Concept of the fuel cycle foresees long term reload period. Within this concept the interval between reloads
equals to 800 effective days while the share of reloaded fuel assemblies is one third. The full campaign at thiscondition is 6.6 years.

11. Waste Management and Disposal Plan
Strategy of the decommissioning and RAW management is being developed during the conclusion of thecontract and is prepared in accordance with the IAEA recommendations.

12. Development Milestones
T0+2
T0+4
T0+6
T0+7
T0+11

70

Development of a preliminary design
Technical requirements, R&D program, basic design
R&D, development of PSAR, expert review and licensing, architecture engineering
Detailed design, fabrication of equipment, construction
Operation license, first criticality, first start, commissioning

KARAT-100 (NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of NIKIET, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type

NIKIET, Russian Federation

Coolant/moderator

Light water / Light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

360 / 100NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

7.0

Number of fuel assemblies in
the core
Fuel enrichment (%)

199

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

45.9

Reactivity control mechanism

Control rods drive mechanism

Approach to safety systems

Passive

Design life (years)

80

Plant footprint (m2)

22 500RPV height/diameter (m)13.25 / 4.00

RPV weight (metric ton)

348

Seismic Design (SSE)

0.3g

Fuel cycle requirements /Approach

Refueling (fuel shuffling) interval is
up to 900 EFPD; Fuel assembly life
cycle is about 7.5 years
Multi-purpose reactor: cogenerationof electricity and heat
Conceptual design

Distinguishing features
Design status

BWR

Natural circulation

104 / 286
UO2 pellet / hexagonal

490

1. Introduction
KARAT-100 is an integral type multi-purpose boiling water reactor (BWR) with a power output of 360 MW(t)and a rated electrical output of 100 MW(e). The design adopts engineering approaches proven at prototype
and testing facilities. The reactor is designed for the production of electrical power, heat for district heatingand hot water in cogeneration mode. The design adopts natural circulation for its primary cooling system coreheat removal in all operational modes. The design configuration incorporates passive safety systems to enhance
the safety and reliability.

2. Target Application
The KARAT-100 is a multipurpose BWR assigned for electricity generation, district heating and forcogeneration. The KARAT-100 power unit has a high load following capability to cope with daily power
variation from 20% to 100% of nominal capacity.

3. Design Philosophy
KARAT-100 reactor is being built as the base reactor for the evolution of power generation in isolated or
remote locations not connected to the unified grid. The key factor that makes this reactor a perfect choice fora nuclear cogeneration plant is its economic competitiveness against other sources of thermal and electric
power, achieved primarily due to a combined generation of heat (for district heating) and electricity.

71

4. Main Design Features
Nuclear Steam Supply System
The reactor uses a single-circuit heat removal system. Steam is generated by the coolant boiling in the reactor
core. The steam discharge system is designed to remove steam from the reactor directly to the turbine plant.
The steam pressure in the system is 7 MPa, at the steam temperature of 286 0C. The humidity of the steam fed
to the turbine plant is 0.1%. KARAT-100 uses gravity-type steam separation with additional drying in louvertype separators without centrifugal axial separators, which improves the stability of the reactor operation.

Reactor Core
The reactor core consists of 199 FAs of proven design. The total number of cells in the core support grid is
253 (there are locations for an extra FA row or the reflector). There are two types of FAs used in the reactor
core: 111 with channels for the CPS rods and 88 without such channels.

Reactivity Control
The reactor core includes 111 control and protection system (CPS) rods. The rod represents a structure
comprising eight absorber elements arranged uniformly in a circumferential direction and retained in a support
grid. The absorber elements are spaced by a spacer grid. The absorber material is boron carbide (B4C). The
rods are accommodated in the FA guide channels and are grouped into clusters of 3 assemblies each to reduce
the number of actuators.

Reactor Pressure Vessel and Internals
KARAT-100 reactor vessel consists of a number of shells, a head and a bottom welded one to another. The
reactor internals - a barrel, a core support grid, a chimney, emergency heat exchangers and the reactor core are accommodated inside the vessel. The reactor vessel has nozzles for feed water supply and steam discharge,
as well as nozzles for the emergency heat exchangers. All nozzles are located in the vessel’s upper part whichguarantees that the necessary volume of coolant is maintained even in the event of a nozzle break.

Reactor Coolant System
The coolant is desalinated light water. The heat removal system is single-circuit. Steam is generated directlyin the reactor vessel and, after being dried in the steam space, is fed to the turbine with a humidity of not morethan 0.1 wt. % using gravity separation and louver-type separators built in the reactor. The steam is dumped
into the condenser downstream of the turbine. Some part of the steam could be extracted for heating of plantsystem’s in house water and for heat supply.

Pressurizer
There is no pressurizer. Pressurization is achieved through negative feedbacks on temperature, power and void
reactivity effect.

5. Safety Features
The major goal of the safety assurance arrangements is to limit the KARAT-100 radiological impacts on the
personnel, the public and the environment during normal operation and in cases of operational occurrencesand emergency events. KARAT-100 safety is ensured through the technological sophistication of design, therequired fabrication, installation, adjustment and testing quality and robustness of the reactor facility’s safety
related systems and components, operating condition diagnostics, quality and timeliness of the equipment
maintenance and repair, in-service monitoring and control of processes, organization of work, and qualificationand discipline of personnel.

Engineered Safety System Approach and Configuration
KARAT-100’s system of engineered and organizational measures forms five defence-in-depth levels:- Conditions for KARAT-100 siting and prevention of anticipated operational occurrences;- Prevention of design-basis accidents by normal operation systems;- Prevention of beyond design-basis accidents by safety systems;- Management of beyond design-basis accidents;- Emergency planning.

Decay Heat Removal System
Residual heat is removed during an accident with a loss of heat removal by normal operation systems with thehelp of coil-type emergency heat exchangers accommodated inside the reactor vessel and emergency coolingtanks. The system is based on a passive principle of action. The coolant is discharged as steam depending on
the decay heat level.

Emergency Core Cooling System
The reactor is cooled down in emergencies caused by a loss of the primary circuit integrity or the reactor power
supply using six independent channels through:72

- The generator coast down;- The passive decay heat removal system, including the emergency heat exchanger;- The emergency cooldown system;- Passive-type water accumulators;- The boron solution injection system;- Cooling the reactor’s metal containment.
Additionally, the power unit with KARAT-100 reactor is equipped with the following safety systems:- A steam localization system downstream of safety valves required to localize radioactive steam release when
the safety valves actuate;- The system for the return of boron solution into the reactor designed to feed the borated coolant back fromthe reactor cavity in the event of a reactor vessel or nozzle break;- The reactor water accumulation system aimed at keeping the water inventory in the accumulators for making
up the reactor in emergencies caused by a decrease in the reactor vessel coolant level;- The emergency power supply system used in the event of a loss of power supply from the energy grid.

Containment System
KARAT-100 reactor is housed in reinforced-concrete containment with a stainless steel liner. The containment
forms an additional barrier against the release of radioactive substances into the environment while limiting at
the same time, by its volume, the coolant loss in the event of reactor vessel break. There are isolation gate
valves installed on pipelines at the containment outlet.

6. Plant Safety and Operational Performances
The major goal of the safety assurance arrangements is to limit KARAT-100 radioactive impacts on the
personnel, the public and the environment during normal operation, and in cases of operational occurrencesand emergency events.
KARAT-100 safety is ensured through the specific transfer and distribution of radioactive substances due towater boiling. The key factors are:
- A high inter-phase barrier (water-steam) to the spreading of nongaseous radionuclides prevents these fromentering the steam-condensate line;- Continuous degassing of coolant and removal of gaseous fission products from the circuit limit theiraccumulation in the circuit.

7. Instrumentation and Control Systems
In-core monitoring allows thermal-hydraulic and neutronic parameters of the reactor core and the in-core
coolant natural circulation circuit to be measured directly and indirectly in different operating modes of thereactor. KARAT-100 reactor is expected to be monitored for neutronic and thermal parameters, including thereactor water level, the core inlet water temperature, and the steam space temperature, and periodically tested
for the water chemical properties.

8. Plant Layout Arrangement
The building layout plan for the land-based power unit of KARAT-100 is designed in such a way that thereactor system, including its servicing systems, spent fuel pool, and auxiliaries are located in double protective
air-crash resistance buildings. The designers also claim that the overall size of the steam generating unit allows
transportation of the reactor by railway.
Ventilation Stack
Reactor Unit
Turbine HallEngineering Support
Building

Vehicle Access

Main Control Room Building

Reactor Building
The reactor unit building performs the function of a primary containment. The reactor unit houses the KARAT100 reactor as well as the systems responsible for the emergency removal of heat from the reactor, theemergency reactor shutdown and the removal of radiolysis products from beneath the reactor head. Besides,

73the reactor building houses an irradiated FA storage facility and its cooling system, as well as the reactor
facility’s handling equipment.

Control Building
The main control room and emergency control room are located in control building adjoining the reactor unit
building, from where the reactor facility is operated and thermal parameters are monitored.

Balance of Plant(i).Turbine Generator Building
The turbine block houses the turbine generator, the steam condensate components and equipment, and a bridge
crane for moving operations. The dimensions of the turbine block are 42 m x 28 m, and its height is 28.4 m.
(ii).Electric Power Systems
The normal power supply system is designed to supply electric power to all plant consumers during normal
operation and anticipated operational snags, including accidents, as well as to deliver the electricity generated
by the turbine plant to offsite and in-house consumers.

9. Design and Licensing Status
At the present time, the KARAT-100 is developed to the level of conceptual design and its further development
is expected to be continued upon the receipt of the commercial request.

10. Fuel Cycle Approach
Concept of the fuel cycle foresees long term reload period. Within this concept the interval between reloads
equals to 900 effective days while the share of reloaded fuel assemblies is one third. The longevity of the fullcampaign at this condition is 7.5 years.

11. Waste Management and Disposal Plan
Strategy of the decommissioning and RAW management is being developed during the conclusion of thecontract and is prepared in accordance with the IAEA recommendations.

12. Development Milestones
T0+1
T0+2
T0+4
T0+6
T0+7
T0+11

74

Development of technical proposal
Development of a preliminary design
Technical requirements, R&D program, basic design
R&D, development of PSAR, expert review and licensing, architecture engineering
Detailed design, fabrication of equipment, construction
Operation license, first criticality, first start, commissioning

RUTA-70 (NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of NIKIET, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type

NIKIET, Russian Federation

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

70 / NA

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

1. Riser shroud
2. Pool metallic liner
3. Core supporting plate with control lead
tubes
4. Reactor core
5. Plenum
6. Check valve
7. Secondary water inlet
8. Secondary water outlet
9. Primary pump
10. Primary HX
11. Upper header
12. Control rod drives
13. Isolation plate

Number of fuel assemblies in
the core
Fuel enrichment (%)

Pool-type

Natural (below 30% of rated
power)/forced (for 30-100%of rated power)
Atmospheric pressure atreactor poll water surface
75 / 102
Cermet (0.6 UO2+ 0.4 Al
alloy) / hexagonal
91
3.0

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

25-30

Reactivity control mechanism

Control rods

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

100 000

RPV height/diameter (m)17.25 / 3.20

Seismic Design (SSE)> 0.8g (automatic shutdown)

Fuel cycle requirements /ApproachDistinguishing features

Once-through fuel cycle with
UO2 fuel
Design capable forradioisotopes production for
medical, neutron beams forneutron therapy and industrial
purposes
Conceptual design

Design status36

1. Introduction
RUTA-70 is a multi-purpose water-cooled water-moderated integral pool-type reactor serving as a Nuclear
Heating Plant (NHP) of 70 MW(t) thermal capacity for district heating, desalination and radioisotopes
production for medical and industrial purposes. It has no power conversion system. In the primary cooling
circuit, the heat from the core is transferred to the primary heat exchanger (HX) by forced convection at fullpower and by natural convection at power below 30% of the rated power. Forced coolant circulation using
pumps for operations at power levels of 30% to 100% rated power increases the coolant flow rate in the primarycircuit and raises the down-comer temperature. RUTA-70 can perform continuous operation without any
maintenance for about one year. RUTA-70 reactors can be located in the immediate vicinity of the heat users.

2. Target Application
The conceptual design of RUTA is primarily developed to provide district heating in remotely isolated areas.
Continuous increase of the organic fuel costs in the country essentially enhances the prospect of RUTA as aheating reactor. In addition, RUTA can also perform seawater and brackish water desalinations based on
distillation process.
75

3. Design Philosophy
The basic design principles of this reactor are design simplicity and a high safety level due to a low pressure
and a large coolant inventory in the primary system. The design aims for low cost of plant construction andoperation, high level of safety achieved through specific features and inherent safety mechanisms. The reactor
facility is a ground based nuclear heating plant (NHP) designed similarly to pool type research reactors.

4. Main Design Features
Nuclear Steam Supply System
RUTA-70 has a two-circuit layout. The primary circuit is an in-pile reactor core cooling circuit and thesecondary circuit is an intermediate one that removes heat from the reactor and transfers it to the third circuit,which is the consumer circuit, i.e., to the heating network. Most of the plant equipment, including the primaryto-secondary side heat exchangers (HX-1/2) resides at dry boxes outside the pool.

Reactor Core
The reactor core is placed in the lower part of the reactor vessel, the vault, in the shell of the chimney section.
The core is designed with the ‘Cermet’ fuel rods that contribute to the reactor safety due to a high thermal
conductivity of the fuel matrix and its role as the additional barrier to the fission products release. The reactor
core consists of 91 fuel assemblies (FA) of hexagonal geometry with 120 fuel rods per each FA. The height ofreactor core is 1400 mm or 1530 mm depending on the fuel rod type. The core equivalent diameter is 1420
mm. In the radial direction, the design of the RUTA-70 fuel assembly is similar to that of the VVER-440 fuelassembly.

Reactivity Control
In the RUTA-70 design, the following mechanisms of reactivity control and power flattening are applied:
optimization of refuelling, use of burnable poison, profiling of fuel loading and movable control rods. The
reactivity control is performed by regulating the control rods and using burnable poison. The reactivity margin
is partly compensated by the burnable absorber (gadolinium) incorporated into the fuel rod matrix in a way to
improve a core power distribution. The rest of the reactivity margin is compensated by the control rod groups.

Reactor Coolant System
The primary coolant forced circulation is provided by two main circulation pumps - one pump per each of two
reactor loops. Two MCPs of axial type are installed in the bypass lines of the main circulation loop close to
the down-comer inlet. The loop arrangement of the primary circuit components, with the secondary circuit
pressure exceeding the pool water pressure, ensures that the reactor coolant is localised within the reactor tank.

Steam Generator
The turbine and associated systems (including steam generator) are not used in the NHP RUTA-70.

Primary Cooling Mechanism
The heat from the core to the primary heat exchanger (HX) is transferred by forced convection of the primarywater coolant at full power operation but by natural convection under operation conditions below 30% of the
rated power. The application of forced coolant circulation using pumps for operations at power levels of 30%
to 100% rated power increases the coolant flow rate in the primary circuit and raise the down-comer
temperature by reducing the water thermal gradient in the reactor core. The distributing header is placed in theupper part of the shell of the chimney section. Pipelines of water supply to the primary HXs are connected tothe header from both sides. Downstream of the HXs, coolant is directed via the suction header to the circulation
pump that supplies water to a group of heat exchangers located at one side of the pool. Water is returned fromthe pump head via the supply header. Pumps are connected to the bypass line of the natural convection circuit
and are placed in a special compartment in close vicinity to the reactor pool.

5. Safety Features
The high safety level of pool reactors is achieved through their design features, which make it possible toresolve some of the major safety issues through the employment of the naturally inherent properties of thereactor. The safety concept of the RUTA-70 is based on the optimum use of inherent safety features, consistent
implementation of defence in depth strategy and to perform the functions based on principles such as multichannelling, redundancy, spatial and functional independence, application of a single failure criterion anddiversity.

Engineered Safety System Approach and Configuration
The RUTA-70 uses mostly passive systems to perform safety functions such as: air heat sink system for
emergency cooldown (ASEC), gravity driven insertion of the control rods in the core as reactor safety control
system, the secondary circuit overpressure protection system, the overpressure protection system for air spacein the reactor pool and pre-stressed concrete external impacts protection system. In case of multiple failures in
the reactivity control systems and devices, safety can be ensured by self-control of reactor power (boiling 76

self-limitation of power), i.e. through the inherent safety features of the reactor. There is a stabilisation ofreactivity feedbacks determined by negative fuel and coolant temperature reactivity coefficients and by the
positive density reactivity coefficient.

Decay Heat Removal System
Natural circulation in the secondary circuit provides for residual heat removal from the shutdown reactor andpassive cooldown of the reactor facility in blackout emergency situation. The passively actuated ASEC
provides residual heat removal to the ultimate heat sink (atmospheric air). ASEC is envisaged for reactor
cooldown in case of loss of auxiliary power. Each loop of the secondary circuit has an ASEC subsystem (train);
the ASEC is connected at the bypass line of the network heat exchangers. If all controlled trains of heat removal
are lost, heat losses via the external surface of the reactor pool to the surrounding environment (ground) areconsidered as an additional safety train. Residual heat is accumulated in the pool water. The transient of poolwater heat-up in the aqueous mode before the onset of boiling takes several days. As soon as boiling starts,
steam goes to the reactor hall where it is condensed by passive condensing facilities. A reactor boil-off without
makeup takes 18 to 20 days. Upon completion of this period residual heat is balanced by heat transfer to theground. Core dry out is avoided. Moderate temperatures are not exceeding the design limits characterize fuelelements.

Emergency Core Cooling System
In emergency situations, residual heat is transferred by natural circulation of the coolant in the reactor tank andin the secondary circuit in station blackout condition. Heat is removed from the secondary circuit convectors
using the ASEC under forced or natural circulation of air in the convector compartments. Direct-acting devices
open air louvers of the ASEC passively. The system for emergency makeup of the primary and secondary
circuits is an active system.

Reactor Pool
The reactor pool consists of reactor core and internals, control and protection system, distributing andcollecting headers and a large amount of water. Big amount of water in the reactor pool ensures slow changing
of coolant parameters and reliable heat transfer from the fuel, even if controlled heat transfer from the reactor
is not available. Fuel temperatures are moderate.

Containment System
The inner surfaces of the pool concrete walls are plated with stainless steel.

6. Plant Safety and Operational Performances
The NHP RUTA-70 may operate in both the base load and load follow modes. Two independent systems basedon diverse drive mechanisms are provided for safe reactor shutdown and ensure the reactor power control. One
system acts as an accident protection system, while the actuated second system is designed to provide
guaranteed sub-criticality for an unlimited period of time and to be able to account for any reactivity effects
including those in accidental states. Either system can operate under the failure of a minimum of one rod withmaximum worth. In case of loss of power to the reactor control and protection system (RCP), all rods of thissystem can be inserted in the core under the effect of gravity.

7. Instrumentation and Control Systems
RCP actuators based on two diverse principles of action have been chosen for the RUTA70:- Multi-position mechanical RCP actuator for automatic (ACR) and manual control rods (MCR);- Two-position hydrodynamic RCP actuator for scram rods (SR).
In the core there are 42 reactor control and protection system (RCP) rods composing two shutdown systems
with diverse actuators. One of these systems intended specifically for core emergency protection (EP) includes
12 rods. The second shutdown system performing the concurrent functions of shutdown and control includes
a group of six (6) automatic regulators (ACRs) and four (4) groups of a total of 24 control rods, for remotemanual reactivity control (manual rods MCRs). In response to the scram signal, all control rods of the second
shutdown system also perform the functions of emergency protection. MCRs are used to compensate forrelatively fast reactivity changes such as heat up and xenon poisoning of the reactor therefore, most of MCRs
will be withdrawn under nominal operating parameters. MCRs and scram rods may take the intermediate
position in the core performing the functions of power control and forming the radial power profile. The slow
transients of reactivity change (such as burn-up of fuel and burnable poison) are also controlled by the groupof ACRs plus the required groups of MCRs.

8. Plant Layout Arrangement
Reactor Building
The protective flooring composed of slabs is installed above the reactor pool to avoid possible damage to theprimary components from external impacts. To prevent gas and vapour penetration to the reactor hall from theupper part of the reactor, joints of the protective slabs are gas-tight.
77

10
9

2 5

1. Core, 2. Primary heat exchanger
3. Check valve, 4. Pump
5. Primary Circuit distributing header
6. Secondary circuit inlet pipeline
7. Secondary circuit outlet pipeline
8. SCS drives, 9. Upper slab
4

6

1

3

8
7

Control Building
The smallest staffing of the operating shift is four persons. These are the NDHP shift supervisor, the chief
reactor control engineer, a fitter-walker for normal operation systems and a duty electrician to attend toelectrical devices and systems, instrumentation and control. A supervising physician and a refuelling operator
are added to the regular shift staff for the core fuelling, first core critical mass attaining, power start up andrefuelling periods. The total personnel number including regular engineers, technicians and administrative staff
may reach up to 40 persons.

Balance of Plant
The turbine and associated systems are not used in the NHP RUTA-70. Such scheme in spite of some reductionof autonomy of district heating system (in comparison to high-temperature reactors) possesses several majoradvantages: Increase in reliability of heat supplying due to diversification of heat sources, provide redundancy
required by relatively cheap heat sources and increase of economic effectiveness of heat production.
1. Reactor pool, 2. Reactor Core
3. Primary heat exchanger, 4. Concrete vessel
5. Soil, 6. Purification system,
7. Ventilation system, 8. Secondary circuit
9. Containment, 10. Residual heat removal system
11. Secondary circuit circulation pump,
12. Secondary circuit pressurizer,
13. Secondary heat exchanger
14. Peak/backup heat source, 15. Control valves
16. Grid circulation pumps, 17. Grid water, 18. Consumers

9. Design and Licensing Status
To provide an operating reference for the reactor, in 2004, the feasibility study was carried out jointly by
NIKIET, IPPE, and Atomenergoproekt (Moscow). This study showed that RUTA-70 could be deployed alongwith the non-nuclear sources of power operating in peak and off-peak mode.

10. Fuel Cycle Approach
The standard fuel cycle option for the RUTA70 NHP is a once-through fuel cycle with uranium dioxide fuel.
The alternative fuel cycle option is a once-through cycle with cermet fuel (microparticles of fuel in a metallic
matrix). Standard fuel reprocessing method as used for VVER type reactors could be applied. In this, fuelreprocessing can be made centralized. According to the design of the NHP RUTA70, spent fuel assemblies
should be stored in the cooling pond for 3 years after discharge from the reactor core and then transported to
the fuel reprocessing plant without further long-term on-site storage.

11. Waste Management and Disposal Plan
Radioactive waste is to be transferred to the ‘National Radioactive Waste Management Operator’ forsubsequent disposal.

12. Development Milestones
1990
1994

2003

78

Conceptual design of the 20 MW(t) RUTA heating plantFeasibility study ‘Underground NHP with 4 × 55 MW(t) RUTA reactors for district
heating in Apatity-city, Murmansk region’
Technical and economic assessments for regional use of the 70 MW(t) RUTA reactor to
improve the district heating system

STAR (STAR ENERGY SA, Switzerland)
All content included in this section has been provided by, and is reproduced with permission of, STAR ENERGY SA, Switzerland.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
STAR ENERGY SA
of origin
Switzerland
Reactor typePressure tube light water reactor
Coolant/moderator
Light water / Light water
Thermal/electrical capacity,
30 / 10MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
12.5 / 4.4Core inlet/outlet coolant
temperature (oC)270 / 300

Fuel type/assembly array

UO2 pellet assemblies in individualpressure channels

Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)

138

Reactivity control mechanism

Control rods and Er / Gd2O3 burnable
absorberPassive60
36 0007.6 / 2.3
20
Degree VIII (MSK-64)
Once-through HALEU; no on-sitestorage
Compact, low-weight design without
a pressure vessel; flexibility ofoperation; intrinsic passive safety

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approachDistinguishing features

Design status

19
120
60

Basic design

1. Introduction“STAR” is a channel-type two-circuit light-water nuclear reactor with a thermal rate of 30 MW(t) and anominal electric power output of 10 MW(e). Based on time-proven solutions utilized in multiple contemporary
light-water reactor designs, “STAR” features a distinctive layout, wherein the coolant of a double-loop primarycircuit circulates through individual pressurized channels instead of an outer pressure vessel, which is the main
factor contributing to the mass and dimensions of traditional PWRs. The channels’ design in the form ofbayonet heat exchangers and the use of UO2 fuel enriched to 19 percent further decrease the reactor unit’s size,
which effectively enables it to be fully assembled in factory and delivered to the installation site by land, sea,or air, substantially decreasing its production and maintenance costs.

2. Target Application
The “STAR” reactor is designed to serve as a versatile tool to fulfil the energy needs of remote or isolated
localities, such as small settlements or power-intensive industrial areas. Depending on the configuration of thesecondary island, it is capable of functioning as a district heating or electric power plant or operate incogeneration mode. Additionally, it can be utilized as a source of heat or steam for water desalination or otherindustrial purposes.

3. Design Philosophy
Since SMRs generally do not enjoy the economy-of-scale benefits offered by larger nuclear power plants, the79main goal of the design is to provide a financially viable alternative to small-scale fossil fuel power stations
while maintaining the required high safety standards. This goal is mainly achieved by combining well-tested
technical solutions used in existing light-water reactors with a unique layout tailored for ease of manufacturing,assembly, transportation, and installation of the reactor. The design assumes that the fabrication process shall
only require commercially available parts and materials to reduce the initial investment required for researchand licensing. It also includes multiple redundant passive safety features following the “defence in depth”
principle with a further goal to reduce the power plant’s total footprint, limit the required controlled zone toits boundaries, and enable deploying it near the end user.

4. Main Design Features
Power Conversion
There are three primary possible variants of nuclear energy conversion at a “STAR” power plant: a nuclear
electric plant referred to as “STAR-E”, a nuclear heating plant (“STAR-T”), and a combined nuclear power
plant (“STAR-C”). The design of the secondary circuits for all three variants is based on the parameters of thereactor unit’s two steam generators with a combined output of 30 MWt. The admission pressure of steam
produced in the steam generator is 4.4 MPa with a steam dryness of 99%. The feed water temperature is 220°C.

Reactor Core
The reactor core consists of 138 pressurechannels, each designed as a bayonet heatexchanger with its outer tube welded into
the lower plate of the primary coolant inlet
chamber and the inner tube welded into thelower plate of the primary coolant outlet
chamber. The coolant is fed to the corethrough the primary coolant inlet chamber,from which it flows along the outer sectionof a pressure channel and rises along theinner tube, which houses a fuel assembly.
The tubes are manufactured from azirconium-niobium alloy, and the space
between the tubes is filled with a neutrontransparent material. The reactor core isencased into a neutron reflector consistingof stacked Al-Be alloy rings with athickness of 100 millimetres.

Reactor core layout

Fuel Characteristics
A “STAR” nuclear fuel assembly consists of 18 nuclear fuel rods placed in a triangular grid in two concentric
circles around a central support rod, fixed in place with spacing meshes. The central support rod is used toattach the fuel assembly to the upper plate of the primary coolant outlet chamber and the drive mechanism.
The fuel rods contain fuel pellets made of uranium dioxide with a maximum enrichment of 19%. Additionally,
6 of the 18 fuel rods contain Er or Gd2O3 burnable absorber, which compensates for loss of reactivity during
normal operation of the reactor and extends its fuel cycle, which is set to 10 years. The nuclear fuel stays inside
the core throughout the entire fuel cycle, including the scheduled maintenance periods.
The inner volume of the nuclear fuel rods is filled with high-purity helium to improve thermal conductivity of
the gas mixture inside the fuel rods and to reduce the average volumetric fuel temperature.

Reactivity Control
Reactivity control is performed with burnable erbium and gadolinium oxide absorbers in a number of the fuelassemblies and 10B reactivity control rods evenly distributed inside the core. Boric acid is not intended forreactivity control during normal operation; however, it can be used for ensuring safe shutdown of the coreduring accident conditions. Additional reactivity control is also provided by the negative thermal feedback of
the moderator.

Reactor Pressure Vessel and Internals
The “STAR” reactor does not feature a pressure vessel. Instead, the pressure of the coolant is contained by theelements of the primary circuit, which, together with the I&C systems, are bundled into a “monoblock” layout
and encased into a sealed cylindrical thin-walled (1 cm thick) steel reactor vessel that ensures convenienttransportation and installation of the reactor unit and serves as an additional containment barrier. The vacant
space between the reactor vessel and the internals is filled with high purity helium. The total height of thereactor module is 7.6 meters with a diameter of 2.3 meters.

Reactor Coolant System
The primary circuit coolant system includes 138 pressure channels containing the reactor core’s fuel80

assemblies, coolant inlet and outlet chambers, and two primary circuit loops, each featuring a pressurizer, amain circulation pump, and a triple-sectioned steam generator. The double-loop layout has been selected toreduce the overall size of the reactor module. The primary circuit is designed to hold an operating pressure of
12.5 MPa and maintain an outlet coolant temperature of 300°C.

Secondary System
The secondary circuit connects to the reactor vessel from the outside through openings for feed water andsteam output pipelines, interfaces with the primary coolant circuit through two steam generator arrays inside
the reactor vessel and provides the final output of the reactor module. Like the primary circuit, the secondary
circuit has forced circulation, so the secondary island must be equipped with a high-pressure feed water pump.
The circuit’s output has a temperature of 256°C with a feed water temperature of 220°C.

Steam Generator
Unlike other reactors currently in operation, the reactor contains two steam generators, each consisting of three
separate sections connected through inlet and outlet pipelines. Each steam generator is designed as a oncethrough unit with an inlet of the primary coolant in its upper part and the outlet in the lower part. The upper
part of a steam generator contains centrifugal and louvered separators. The combined thermal power rate of all
the steam generators is 30 MW.

5. Safety Features
Engineered Safety System Approach and Configuration
The design of the “STAR” reactor follows the principle of inherent safety and provides for both long-term heatremoval after reactor shutdown and emergency cooling during design basis accidents, including mechanicallytriggered safety measures requiring no power source or human input.

Decay Heat Removal System / Reactor Cooling Philosophy
During normal operation, sufficient radiation and thermal protection is achieved with several layers ofshielding, including the reactor core’s reflector, a cushion of pure inert gas, an airtight reactor vessel, and areactor shaft shielding. In the event of an emergency reactor shutdown, the design of “STAR” offers systems
for both short-term emergency cooling and long-term residual heat removal. The short-term emergency coolingsystem is triggered in a loss of coolant accident by several mutually independent mechanical sensors. The
system injects pressurized water from an emergency tank into the reactor vessel to provide cooling to the coreand prevent its exposure. The long-term cooling system is designed to passively remove decay heat from thefirst circuit through steam generators; it has been estimated that the first circuit shall maintain natural
circulation of coolant even in the event of failure of both main circulation pumps. Finally, in the occurrence ofa guillotine rupture of a primary circuit pipeline, additional cooling shall be provided by a steam condenser
installed at the top of the reactor vessel.

Spent Fuel Cooling Safety Approach / System
Initial calculations and modelling show that, given proper handling and fractioning, spent fuel from a “STAR”
reactor cools passively and does not require temporary on-site storage in a spent fuel pool. Requirements toequipment and procedures for waste management shall be elaborated at further stages of the design.

Containment System
The design’s containment system is based on the “defence in depth” philosophy and includes a set of technicaland administrative measures aimed at preventing environmental exposure to radiation and radioactive
substances from the reactor.
The set of barriers includes:- Fuel pellet matrix;- Fuel rod casings;- Pipelines end equipment of primary coolant circuit;- Reactor core pressure channels;- Reactor core shielding;- Airtight reactor vessel;- Ferrocement reactor shaft.

Chemical Control
The functions of and requirements for the chemical control systems shall be elaborated on further stages of thedesign.

6. Plant Safety and Operational Performances
From the beginning of development of “STAR”, reasonable safety requirements to be imposed for the reactor’s
design features are as follows:- Core damage frequency – less than 10-6/reactor year
- Large early radioactivity release frequency – less than 10-7/reactor year81

7. Instrumentation and Control System
On the conceptual level, the I&C system isdivided into the following main blocks:- Primary and reserve control rooms;- Reactor control and protection system,
including remote controls and two independent
emergency shutdown systems;- Process parameters control system;- Internal reactor control system;- Normal operation control systems;- Miscellaneous automatic systems.

8. Plant Layout Arrangement
The default design suggests deploying thereactor unit in an underground ferrocement
reactor shaft to improve its containment
capabilities, enhance resilience to external
threats, and increase convenience of emergency
water tank placement. However, other reactor
building arrangements are also available.
Secondary island and civil part buildings should
be designed with consideration for the given
country’s legislation and environment.

9. Testing Conducted for
Verification and Validation“STAR” module placement option
inside an underground reactor shaft

DesignMost parts and materials used for the “STAR”
design are tested, commercially available andhave been nationally or internationally certified.
During the conceptual design stage, extensive
physical calculations and modelling for thereactor unit were completed using MCNP, Example layout of a reactor building with an adjacent secondary island
WIMSD4, and SSL DYNCO codes. Further
means of validation, including probabilistic and deterministic safety analysis tools and live test facilities forvarious components, shall be developed on the later design stages.

10. Design and Licensing Status
Interaction with Nuclear Safety Authority – Started; Site permit for FOK plant – To be developed.

11. Fuel Cycle Approach
The design assumes six fuel loading throughout the reactor’s lifetime: the initial loading and 5 reloads every
10 years. Apart from description and requirements for refuelling equipment and procedures to be developedon later design stages, current regulations and best practices concerning handling of HALEU fuel areapplicable to this design.

12. Waste Management and Disposal Plan
By design, the “STAR” reactor facility does not release liquid or solid radioactive waste during normal
operation. Any solid or liquid radioactive waste collected during refuelling and maintenance shall be disposedof in accordance with applicable regulations and practices.

13. Development Milestones
2015
2019
2021
Current (2022)

2024
2025
2026-2027

82

Initial concept developed; patent priority date secured
Concept submitted for peer review
Conceptual design developed; preliminary feasibility study for a FOAK unit completed
Conduction of feasibility study for FOAK unit and initiation of licensing procedures withlocal regulator and basic design
Development of basic design and PSAR, conduction of site survey
Launch of FOAK plant project, development of detailed design and extended SAR
Pilot unit installation and launchRolls-Royce SMR (Rolls-Royce SMR Ltd,
United Kingdom)
All content included in this section has been provided by, and is reproduced with permission of Rolls-Royce SMR Ltd, United Kingdom.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country ofRolls-Royce SMR
originKingdom
Reactor type3-loop PWR
Coolant/moderator
Light water

Ltd,

UnitedThermal/electrical capacity,
MW(t)/MW(e)1 358 / 470

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15.5 / 7.8

Core inlet/outlet coolant
temperature (oC)

295 / 325

Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup (GWd/ton)Reactivity control mechanism

UO2 / 17 × 17 Square
121

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements/approachDistinguishing features

Passive and Active
60
40 0007.9 / 4.2150
>0.3g
Open cycle
Modular approach facilitating rapid
and cost-effective build.
Detailed design

Design status

<4.95
18
50 – 60
Control rods

1. Introduction
The Rolls-Royce SMR has been developed to deliver a market driven, affordable, low carbon, energy
generation capability. The developed design is based on optimised and enhanced use of proven technologies
that presents a class leading safety outlook and attractive market offering with minimum regulatory risk. Rapid,
certain and repeatable build is enhanced through site layout optimisation and maximising modular build,standardisation and commoditisation.

2. Target Application
The Rolls-Royce SMR is primarily intended to supply baseload electricity for both coastal and inland siting.
The design can be configured to support other heat-requiring or cogeneration applications, as well as provide
a primary, carbon free, power source for the production of e-fuels.

3. Design Philosophy
The design philosophy for the Rolls-Royce SMR is to optimise levelised cost of electricity with a low capitalcost. The power output is maximised whilst delivering robust economics for nuclear power plant investment
and a plant size that enables modularisation and standardisation throughout.

4. Main Design Features
Nuclear Steam Supply System
The Rolls-Royce SMR is a close-coupled three-loop Pressurised Water Reactor employing an indirect Rankine
cycle. Coolant is circulated via three centrifugal Reactor Coolant Pumps to three corresponding vertical u-tube

83

Steam Generators.

Reactor Core
The nuclear fuel is industry standard UO2 enriched up to 4.95%, clad with a zirconium alloy and arranged ina 17x17 assembly. The core contains 121 fuel assemblies and has an active fuelled length of 2.8 m, deliveringa thermal power of 1358 MW(t). Through cycle reactivity compensation is provided using gadolinia neutron
poison. Different fuel assemblies contain different poison loadings, through inclusion of different numbers of
poison pins and different gadolinia weight loading. All poison pins in a single fuel assembly have the samegadolinia loading.

Reactivity Control
Duty reactivity control is provided through movement of control rods and use of the negative moderator
temperature coefficient inherent to Pressurised Water Reactors. No concentration of soluble boron is
maintained in the primary coolant for duty reactivity control, which facilitates a simplified plant design andeliminates risks associated with boric acid as well as the environmental impact of boron discharge.

Reactor Pressure Vessel and Internals
The Reactor Pressure Vessel comprises the upper shell containing the six Reactor Coolant Loop nozzles, three
Direct Vessel Injection nozzles, the lower shell consisting of a plain cylindrical forging with no circumferential
welds present within the high flux region of the core and the torispherical lower head.
The Integrated Head Package forms an assembly of the reactor head area components. It comprises the Reactor
Pressure Vessel Closure Head with torispherical profile and outer bolting flange, Control Rod Drive
Mechanisms and their cooling system, In-Core Instrumentation, integral missile shield and lifting assembly.
The lower Reactor Pressure Vessel Internals comprise the core barrel which holds the fuel and diverts coolant
flow from the coolant inlets to the flow distribution device, the flow distribution device which diverts,
straightens and distributes the coolant flow prior to entry into the inlet plenum and the radial neutron reflector.
The upper Reactor Pressure Vessel Internals take coolant from the outlet of the fuel assemblies and transfer itto the outlets. The upper internals comprise the upper support barrel which holds the fuel in place and thecontrol rod housing columns which align and support the control rods and control rod drive shafts.

Reactor Coolant System
The Reactor Coolant System is a three-loop close-coupled configuration.
Steam Generators are located around the circumference of the Reactor
Pressure Vessel, with short close-coupled pipework connections between
them. The pressuriser is connected to the pipework hot leg. A centrifugal
Reactor Coolant Pump is mounted directly to the outlet nozzle of thecorresponding Steam Generator. Each Steam Generator is elevated above
the Reactor Pressure Vessel to ensure that a sufficient thermal driving
head is available for natural circulation flow for situations where pumped
flow from the Reactor Coolant Pumps is unavailable.

Steam Generator
A vertical u-tube Steam Generator design has been selected as a matureand readily deployable technology. The Steam Generators include an
integral crossflow preheater which works by preferentially directing
feedwater to the cold side of the tube bundle, resulting in an increased
thermal efficiency and smaller component compared to a non-preheater
design.

Reactor Coolant System

Pressuriser
The primary circuit pressure is controlled by electrical heaters located at the base of the pressuriser and sprayfrom a nozzle located at the top. Steam and water are maintained in equilibrium to provide the necessary
overpressure. The pressuriser is a vertical, cylindrical vessel constructed from low alloy steel, sized to provide
passive fault response for bounding faults, with rapid and significant cooldown or heat-up accommodated.

Primary pumps
Forced Reactor Coolant System flow is provided via three Reactor Coolant Pumps utilising a seal-less design.
Pumped flow supports power operation with Reactor Coolant Pumps also configured to enable natural
circulation flow during conditions where pumped flow is unavailable, ensuring the availability of passive heatrejection.

5. Safety Features
Engineered Safety System Approach and Configuration
The Rolls-Royce SMR design has been developed through a combined system engineering and safetyassessment approach. The safety informed design supports the process by which risks are demonstrated to be84

acceptable and as low as reasonably practicable. Defence in depth is provided through multiple layers of fault
prevention and protection in the form of independent and diverse active and passive systems, with multiple
trains per system. Passive safety systems are designed to deliver their safety functionality autonomously for
72 hours, minimising the demand on human actions and electrical power.

Decay Heat Removal System
The Rolls-Royce SMR employs both active and passive decay heat removal systems. Condenser Decay Heat
Removal utilises the Steam Generators and the normal duty steam condenser to cool the primary plant. It
utilises most of the same equipment used for steam condensing during power operation. Heat is eventually
passed to the tertiary heat sink. Passive Decay Heat Removal is a dedicated safety system that utilises the
Steam Generators and the Local Ultimate Heatsink system to cool the primary plant. Both means of heatrejection are closed loop between the Reactor Pressure Vessel and the Steam Generators and decay heat can bepassed from the core to the Steam Generator through either pumped reactor circuit flow or natural circulation
flow.

Emergency Core Cooling System
Emergency Core Cooling provides decay heat removal through depressurisation of the Reactor Coolant Systemand sustained supply of injected coolant by gravity feed. The function includes three accumulators connected
directly to the Reactor Vessel pipework associated with gravity drain from the Refuelling Pool to the Reactor
Vessel and three containment sump screens. The Emergency Core Cooling function also places requirementson the Reactor Coolant Pressure Relief System for emergency blowdown and Local Ultimate Heatsink System
to transfer heat from the Passive Containment Cooling heat exchangers to the environment.

Spent Fuel Cooling Safety Approach / System
Active and passive cooling systems provide robust decay heat removal defence in depth during fault
conditions. Reactivity control is provided in a soluble boron-free environment using geometric spacing andsolid neutron poison racks.

Containment System
The reactor circuit and other key systems are located within a steel containment vessel to confine release ofmaterials during faulted and accident conditions. The Rolls-Royce SMR also adopts in-vessel retention toconfine the postulated melt in severe accidents.

6. Plant Safety and Operational Performances
Plant conditions have been analysed using industry validated codes to demonstrate significant safety margin
across the levels of defence in depth. The Probabilistic Safety Assessment calculates a core damage frequency
from plant faults of <1E-07 per year of power operation, and that no single fault or class of faults makes adisproportionate contribution to risk, i.e. a balanced risk profile is achieved. Internal and external hazard
assessments have defined the design basis and informed the plant layout to optimise segregation. Key
equipment is protected by the hazard shield which is resilient against external hazards including aircraft impact
and tsunami.
The Rolls-Royce SMR is designed for full compliance with the U.K. Grid Code, including frequent load
following between 50% and 100% power at a rate of 3-5% per minute with a design target to allow operation
at lower power e.g. house load operation (stable operation for at least two hours supplying only the power
station’s house load).
A 15-day outage period is targeted to support high availability (at least 90% availability over the 60-year lifeof the Rolls-Royce SMR), including a target period of 8 days for refuelling.

7. Instrumentation and Control System
The Rolls-Royce SMR plant is controlled and protected by a number of control and instrumentation systems.
The Reactor Plant Control System manages duty operations and uses an available in industry programmable
logic controller or distributed control system. It uses mixed analogue and non-programmable digital sensors
and communicates on hardwired multichannel digital electrical networks. Opportunities to use smart devices
and wireless technologies are being pursued.
The Reactor Protection System provides shutdown in response to a fault. The Reactor Protection System
contains priority logic, which from the range of input signals received determines whether to initiate reactor
shutdown. The Hardwired Diverse Protection System uses non-programmable electronics and as such provides
a diverse means to shut down the plant in response to fault conditions.
Post-Accident and Severe Accident Management Systems within the Nuclear C&I System provide clear plantstatus displays, over the days and months following a postulated accident.

85

8. Plant Layout Arrangement
The power station is designed for installation on an extensive range of inland and coastal sites, across a wide
range of soil and earth conditions, whilst maintaining a compact site footprint of approximately 40,000 m2.
This flexibility is enabled through design features such as seismic isolation for key safety areas. The nuclear
reactor is located in the Reactor Island, adjacent to Turbine Island with the Cooling Water Island following.
Support buildings and auxiliary services are situated within a berm that sweeps around the site and provides alayer of protection from external hazards.

Plant Layout

9. Testing Conducted for Design Verification and Validation
The Rolls-Royce SMR design is based on optimised and enhanced use of proven technologies. Test rigs havebeen designed for validation of safety claims of the key safety systems, utilising established test facilities
reflecting that safety systems are based on readily provable phenomena.

10. Design and Licensing Status
The Rolls-Royce SMR entered formal design assessment by UK regulators in 2022 and targets completion intime for construction of the first of a kind power station to commence in 2026.

11. Fuel Cycle Approach
The Rolls-Royce SMR operates on an 18-month fuel cycle, with a three-batch equilibrium core. Refuelling ismanaged through the provision of an in-containment refuelling pool which temporarily stores both new andused fuel during a refuelling outage. Used fuel is subsequently transferred to an external spent fuel pool forstorage prior to transfer to long term dry cask storage.

12. Waste Management and Disposal Plan
The Rolls-Royce SMR waste treatment systems are based on use of proven technologies and best available
techniques. Industry lessons learned and good practices have been used in the development of systems tominimise active and non-active wastes and discharges, through both design and operational practices adopted.
Standardised waste treatment system components and modules are used to achieve the flexibility required forthe waste informed design. Operation without soluble boron in the primary coolant allows significant reduction
in environmental discharges and concomitant simplification of the waste treatment systems.

13. Development Milestones
2015
2016
2017
2022
2026
2030

86

Rolls-Royce development of initial reference design
Formation of consortium for design of whole power station concept
Mature design concept developed
Formal regulation entered in the UK
Projected start of first of a kind construction
Planned first of a kind commercial operation

VOYGRTM (NuScale Power Corporation,
United States of America)
All content included in this section has been provided by, and is reproduced with permission of, NuScale Power LLC., USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

NuScale Power Corporation, USA

Coolant/moderator

Light water / Light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

250 / 77 (gross)Integral PWR

Natural circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

13.8 / 4.3

Number of fuel assemblies in thecore
Fuel enrichment (%)

37

249 / 316
UO2 pellet / 17x17 square≤ 4.95

Core Discharge Burnup (GWd/ton)≥ 45

Refuelling Cycle (months)

Nominal 18

Reactivity control mechanism

Control rod drive, boron

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

140 000 (VOYGRTM-12)

RPV height/diameter (m)17.7 / 2.7

RPV weight (metric ton)

TBC

Seismic Design (SSE)

0.5g

Fuel cycle requirements /ApproachDistinguishing features

Nominal three-stage in-out
refuelling scheme
Unlimited time for core coolingwithout AC or DC power, or
water addition, or operator action
Equipment Manufacturing in
Progress

Design status

1. Introduction
The NuScale Power Module™ (NPM) is a small, light-water-cooled pressurized-water reactor (PWR).
NuScale VOYGR™ SMR plants are scalable and can be built to accommodate a varying number of NPMs tomeet customer’s energy demands. NuScale standard plants include the VOYGR-4 at 308 MW(e), the VOYGR6 at 462 MW(e), and the VOYGR-12 at 924 MW(e). A six-module configuration is the reference plant sizefor the Standard Plant Design Approval design and licensing activities. Each NPM is a self-contained modulethat operates independently of the other modules in a multi-module configuration. All modules are managed
from a single control room.

2. Target Application
The NuScale design is a modular reactor for electricity production, with the capability for flexible operations
to load follow, and for non-electrical process heat applications, including the cogeneration of heat andelectricity.

3. Design Philosophy
The NuScale VOYGR SMR plant design adopts design simplification, proven light-water reactor technology,
modular nuclear steam supply system, factory-fabricated power modules, and passive safety systems that allow
for unlimited reactor cooling time after a beyond design basis accident without AC or DC power, operator
action, or makeup water. There are no design basis accidents that uncover the core or require operator action.
87

The NPM is designed to operate efficiently at full-power conditions using natural circulation as the means ofproviding core coolant flow, eliminating the need for reactor coolant pumps.

4. Main Design Features
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) consists of a reactor core, helical coil steam generators, and apressurizer within a reactor pressure vessel (RPV). The NSSS is enclosed in an approximately cylindricalcontainment vessel (CNV) that sits in the reactor pool structure. Each power module is connected to a dedicatedturbine-generator unit and balance-of-plant systems.

Reactor Core
The core configuration for the NPM consists of 37 fuel assemblies and 16 control rod assemblies. The fuelassembly design is an approved, commercially available, 17 x 17 PWR fuel assembly with 24 guide tube
locations for control rod fingers and a central instrument tube. The assembly is nominally half the height ofstandard plant fuel and is supported by five spacer grids. The fuel is UO2 with Gd2O3 as a burnable absorberhomogeneously mixed within the fuel for select rod locations. The 235U enrichment is up to the current U.S.
manufacturer limit of 4.95 percent enrichment.

Reactivity Control
Reactivity control in each NPM is achieved mainly through soluble boron in the primary coolant and 16 control
rod assemblies. The control rods are organized into two groups: a control group and a shutdown group. The
control group, consisting of four rods symmetrically located in the core, functions as a regulating group that isused during normal plant operation to control reactivity. The shutdown group comprising 12 rods is used during
shutdown and scram events. The absorber material in these control rods is B4C and their length is 2 m.

Reactor Pressure Vessel and Internals
The RPV consists of a cylindrical steel vessel with an inside diameter of 2.7 m, an overall height ofapproximately 17.7 m, and is designed for an operating pressure of 13.8 MPa. The upper and lower heads aretorispherical and the lower portion of the vessel has flanges just above the core region to provide access forrefuelling. The RPV upper head supports the control rod drive mechanisms. Nozzles on the upper head provide
connections for the reactor safety valves, the reactor vent valves, and the secondary system steam piping.

Reactor Coolant System and Steam Generator
The reactor coolant system (RCS) provides for the circulation of the primary coolant using natural circulation.
Hence, the RCS does not require reactor coolant pumps or an external piping system to generate operationalflows. The RCS includes the reactor pressure vessel (RPV) and integral pressurizer, the reactor vessel internals,
the reactor safety valves, RCS piping inside the containment vessel, and others. Each NPM uses two interwoven, once-through helical-coil steam generators for steam production. The steam generators are located in
the annular space between the hot leg riser and the RPV inside diameter wall.

Pressurizer
The internal pressurizer provides the primary means for controlling reactor coolant system pressure. It is
designed to maintain a constant reactor coolant pressure during operation. Reactor coolant pressure is increased
by applying power to a bank of heaters installed above the pressurizer baffle plate. Pressure is reduced using
sprays provided by the chemical and volume control system (CVCS).

5. Safety Features
The NuScale VOYGR SMR plant adopts a set of engineered safety features designed to provide reliable longterm core cooling under all conditions, including severe accident mitigation. They include integral primarysystem configuration, a containment vessel, passive heat removal systems, and severe accident mitigation
features. This fully passive safety design is rigorously proven by the NuScale Triple Crown for Nuclear Plant
Safety™, which ensures that reactors will safely shut down and self-cool, indefinitely, and do so with no needfor operator or computer action, AC or DC power, or the addition of water—a first for light water reactor
technology.

Engineered Safety System Approach and Configuration
Each NPM incorporates several simple, redundant, and independent safety features, which are discussed asfollows.

88

Cut-away view of NuScale power plant

Site Layout of 12 module VOYGR-12 plant

Safety Approach and Configuration to Manage DBC
The decay heat removal system (DHRS) provides secondary side reactor cooling for non-LOCA events whennormal feedwater is not available. The system is a closed-loop, two-phase natural circulation cooling system.
Two trains of decay heat removal equipment are provided, one attached to each steam generator loop. Eachtrain is capable of removing 100% of the decay heat load and cooling the primary coolant system. Each train
has a passive condenser immersed in the reactor pool.

Safety Approach and Configuration to Manage DEC
Emergency core cooling system (ECCS) consists of three independent reactor vent valves (RVVs) and two
independent reactor recirculation valves (RRVs). The ECCS provides decay heat removal in the unlikely eventof a loss of feedwater flow, combined with the loss of both trains of the DHRS. The ECCS removes heat andlimits containment pressure by steam condensation on, and convective heat transfer to, the inside surface of
the CNV.

Containment System
The functions of CNV are to contain the release of radioactivity following postulated accidents, protect the
RPV from external hazards, and to provide heat rejection to the reactor pool following ECCS actuation. Each
CNV consists of a steel cylinder with an outside diameter of 4.5 m and an overall height of 23.1 m. The CNV
houses the RPV, control rod drive mechanisms, and associated piping and components of the NSSS. The CNV
is immersed in the reactor pool, which provides an assured passive heat sink for containment heat removal
under LOCA conditions.

Spent Fuel Cooling Safety Approach
The used fuel pool provides storage for up to 10 years of used fuel storage, plus temporary storage for new
fuel assemblies. The pool is connected to the ultimate heat sink, and hence, protected by the reactor building.
The pool water volume provides approximately 150 days of passive cooling of the used fuel assemblies
following a loss of all electrical power without the need for additional water. The VOYGR plant site layout
includes space allocation adequate for the dry storage of all the used fuel for the 60-year life of the plant.

6. Plant Safety and Operational Performances
Each NPM is operated independent of other modules. A module is refuelled by disconnecting it from its
operations bay and moving it to a common refuelling area within the shared reactor pool. The module isdisassembled into three major components. After inspecting the module sections and refuelling the core, themodule is reassembled and moved to its operations bay and reconnected to steam and feedwater lines. Other
modules in the plant continue to operate while one module is refuelled. The nominal plant capacity factor is
95% with a refuelling outage time of 10 days.

7. Instrumentation and Control System
The NuScale design includes a fully digital control system based on the use of field programmable gate array(FPGA) technology. The highly integrated protection system (HIPS) platform, approved by the NRC, is basedon the fundamental I&C design principles of independence, redundancy, predictability, repeatability, anddiversity. The HIPS platform is comprised of four module types that can be interconnected to implement
multiple configurations to support various types of reactor safety systems. The FPGA technology is notvulnerable to cyber-attacks.

8. Plant Layout Arrangement
Reactor Building
The NuScale VOYGR SMR plant consists primarily of a reactor building, a control room building, two turbinegenerator buildings, a radwaste treatment building, forced-draft cooling towers, a switchyard, and a dry-cast89

storage area for discharged fuel. The reactor building consists of up to 12 NPMs, module assembly/disassembly
equipment, fuel handling equipment, and a spent fuel pool. Each NPM operates immersed within a common
reactor pool in a separate bay with a concrete cover that serves as a biological shield.

Control Building
The main control room is housed in the control building located adjacent to the reactor building. All NPMs arecontrolled from a single control room. The NRC has approved the control room conduct of operations for 3
operators controlling up to 12 NuScale reactors and has eliminated the requirement for a Shift TechnicalAdvisor.

Balance of Plant
A NuScale VOYGR SMR plant has two separate turbine buildings, each housing up to six turbines and aircooled generators. Each turbine-generator is associated with a single NPM and has dedicated condensate andfeedwater pumps.

9. Testing Conducted for Design Verification and Validation
NuScale has designed, built, and operates a one-third scale prototype testing facility that is used to test the
NPM and gather data for thermal hydraulic codes, safety analysis code, and reactor design validation. NuScale
has designed and performed comprehensive testing to validate the operation of the helical coil steam generators,
the safety valves, the control rod assembly and drive shafts, fuel and various other systems. NuScale has anaverage Technology Readiness Level (TRL) of 8. NuScale testing programs have been audited by the NRC.

10. Design and Licensing Status
In December 2016, NuScale Power submitted the Design Certification Application (DCA) to the NRC. In
September 2020, the NRC issued the Standard Design Approval for the NuScale DCA, making it the first ever
SMR to receive NRC design approval. NuScale is applying for a power uprate to 250 MWt in 2022, with an
expected NRC review completion in 2024. The first plant will be a 6-module VOYGR-6 plant located near
Idaho Falls, Idaho with owner, Utah Associated Municipal Power Systems (UAMPS). The target commercial
operation date for the first plant is 2029.

11. Fuel Cycle Approach
The three-batch refuelling is conducted on a nominal 18-month refuelling cycle in an “in-out” shuffle scheme.
During the refuelling process, one-third of the fuel assemblies are removed from the NPM and placed in the
spent fuel pool. Actual batch size, loading pattern, and cycle length are customer driven optimization
requirements.

12. Waste Management and Disposal Plan
Removed assemblies are stored in the used fuel pool for initial cool-down and later moved to an on-site drycask storage. The plant design includes sufficient on-site storage space for all the spent fuel produced during
the 60-year life of the plant. Final disposal is expected to be in a national fuel repository when available.

13. Staffing
A VOYGR-12 plant will require a minimum of 3 licensed operators per shift in the control room and isestimated to require 270 plant employees for normal operation and maintenance. No Shift Technical Advisor
is required in the U.S. for this design.

14. Development Milestones
2003
2007
2008
2011
2013
2016
2020
2022
2023
2025
2029

90

Initial concept developed and integral test facility operational
NuScale Power, LLC was formed
Initiation of NRC Pre-Application
NuScale Power is acquired by Fluor
U.S. DOE SMR Cooperative Agreement signed
Design certification application was submitted to the U.S. NRC
Standard Design Approval (SDA) received from the NRC in September 2020
NPM production began with the production of forging dies for the Upper RPV
Combined License Application (COLA) to be submitted to the NRC for first plant
Construction targeted to start for first VOYGR-6 plant in Idaho Fall, Idaho, USA
First commercial NuScale VOYGR-6 plant targeted to be operational in Idaho

BWRX-300 (GE-Hitachi Nuclear Energy,
USA and Hitachi-GE Nuclear Energy, Japan)
All content included in this section has been provided by, and is reproduced with permission of GE-Hitachi Nuclear Energy, USA.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
GE-Hitachi Nuclear Energy, United
country of origin
States and Hitachi-GE Nuclear Energy,
Japan
Reactor type
Boiling water reactor
Coolant/moderator
Light-water / light-water
Thermal/electrical capacity,
870 / 270 – 290
MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa
Core inlet/outlet coolant
temperature (oC)
Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)
Reactivity control
mechanismApproach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approachDistinguishing features
Design status

Natural circulation-7.2 / 7.2/ Direct
Directcyclecycle
270 / 288
UO2 / 10×10 array-240

3.81 (avg) / 4.95 (max)
12 - 24
49.6
Rods and solid burnable absorber (B4C,
Hf, Gd2O3)
Fully passive
60
9 800
26 / 4485
0.3g
Open fuel cycle utilizing standard BWR
fuel
Natural circulation BWR, integral RPV
isolation valves, isolation condenser
Detailed design

1. Introduction
GE-Hitachi Nuclear Energy’s (GEH’s) BWRX-300 is a designed-to-cost 300 MWe water-cooled natural
circulation small modular reactor (SMR) utilizing simple natural phenomena-driven safety systems. It is thetenth generation of the Boiling Water Reactor (BWR) and represents the simplest BWR design since GE began
developing nuclear reactors in 1955. BWRX-300 is an evolution of the U.S. Nuclear Regulatory Commission(NRC) licensed 1,520 MWe Economic Simplified Boiling Water Reactor (ESBWR).

2. Target Application
Target applications include base load electricity generation, load following electrical generation within a rangeof 50 to 100% power, district heating, synthetic fuel production, hydrogen production and other process heatapplications.

3. Design Philosophy
The BWRX-300 leverages the U.S. NRC approved ESBWR design, proven in-use materials, off-the-shelfcomponents, and operational parameters within the range of the existing BWR experience base. It uses a welldefined, robust, procedure driven process to direct cost-effective design decisions. The BWRX-300 design isnear term deployable with the lead unit planning to enter commercial operation in 2028. It utilizes a coolant
conservation strategy to mitigate loss of coolant accidents which results in significantly simplified structuresand systems with improved safety.

91

4. Main Design Features
Nuclear Steam Supply System
The BWRX-300 Nuclear Steam Supply System (NSSS) leverages the proven supply chain of the ABWR andthe natural phenomena driven safety features from the ESBWR. Components in this system include the Reactor
Pressure Vessel (RPV) and fine motion control rod drives (FMCRDs). The NSSS delivers steam from the RPV
to the turbine via the main steam system; and the condensate and feedwater system delivers feedwater to the
RPV. It also provides overpressure protection of the reactor coolant pressure boundary (RCPB).

Reactor Core
The reactor core of the BWRX-300 is arranged as an upright cylinder containing fuel assemblies located within
the core shroud. During operation, the coolant/moderator enters the bottom of the core as subcooled water andexists the core as saturated steam. The BWRX-300 utilizes GNF2 fuel which is the same fuel as used by the
majority of the BWR fleet today. GNF2 fuel design is a 10x10 array of 78 full-length fuel rods, 14 part-lengthrods and two large central water rods.

Reactivity Control
Reactivity control is provided by control rods loaded with either B4C or Hf neutron absorbers and burnableneutron absorber loaded in the fuel rods. The control rods are moved using the FMCRDs that are successfully
deployed in the ABWR. FMCRDs have two independent means of moving the control rods–motor driven finecontrol for reactivity control during normal operation and hydraulic rapid insertion during a scram.

Reactor Pressure Vessel and Internals
The BWRX-300’s RPV assembly consists of a pressure vessel with removable head, internal components andappurtenances. The reactor internal include the core (control rods and nuclear instrumentation), core support
and alignment structures (shroud, shroud support, top guide, core plate, control rod guide tube and orificed
fuel support), chimney, chimney head and steam separator assembly and steam dryer assembly.

Reactor Coolant System
The reactor coolant system (RCS) is natural circulation driven and provides cooling of the reactor core in all
operational states and all postulated off normal conditions. The BWRX-300 leverages natural circulation
modelling and operational information from the ESBWR and the Dodewaard Nuclear Power Plant that
operated in the Netherlands.

Steam Generator
The BWRX-300 is a direct cycle plant with steam generated directly from the RPV. The RPV has internal
steam separators and driers to provide high quality steam to the steam turbine. Steam generators are notrequired.

Primary pumps
The BWRX-300 utilizes natural circulation and requires no primary pumps.

5. Safety Features

Defense in Depth

92

Engineered Safety System Approach and Configuration
The BWRX-300 is designed using IAEA guidelines to streamline licensing in many countries. The basic
BWRX-300 safety design philosophy is built on utilization of inherent margins (e.g., larger structure volumes
and water inventory) to eliminate system challenges and can easily accommodate transients. Natural driven
phenomena safety-related systems mitigate accidents without the need for AC power. The BWRX-300 defence
in depth concept uses Fundamental Safety Functions (FSF) to define the interface between the defence lines
and the physical barriers.

Decay Heat Removal System
The normal, non-Safety Class 1, shutdown cooling (SDC) system for decay heat removal comprises two
independent pump and heat exchanger trains. These trains provide redundant decay heat cooling capacity. The
major components of each train are a pump and a HX, along with valves, piping, instrumentation and controls,
and power inputs.

Emergency Core Cooling System
The ECCS for the BWRX-300 is the Isolation
Condenser System (ICS). The ICS removes decay heat
after any reactor isolation and shutdown from power
operations under normal or abnormal conditions. The
ICS consists of three independent loops eachcontaining a heat exchanger (HX) and isolation
condenser (IC) with capacity of approximately 33 MW.
The ICS tubes condense steam from the RPV on theinside surface and transfer heat on the outside surface
to the IC pool which is at atmospheric pressure. This
steam condensation and gravity allow BWRX-300 tocool itself for a minimum of seven days without poweror operator action.

Spent Fuel Cooling Safety Approach / System

Isolation Condenser

The Fuel Pool Cooling and Cleanup System (FPC) provides cooling, filtration, and demineralization of thewater in the spent fuel pool (SFP), reactor well, and equipment pool. The FPC consists of two parallel trainsof equipment. Each train includes a pump and heat exchanger. Water is passed from the pump through a filter
element and mixed resin deep bed demineralizer, then passed through a heat exchanger to remove thermalenergy before being returned to the pools.

Containment System
The BWRX-300 Primary Containment Vessel (PCV) encloses the RPV, provides radiation shielding and acts
as a boundary for radioactive released from the RPV. The PCV is a vertical cylinder approximately 16 meters
diameter and 44 meters high. It is integral to and surrounded by the Reactor Building (RB). The PCV is dry
and is located mostly below grade.

6. Plant Safety and Operational Performances
The defense in depth approach utilized in the BWRX-300 results in an internal event core damage frequency
less than 10-7 per year and a large release frequency less than 10-7 per year. The lessons learned and bestpractices of the BWR fleet have been applied to the BWRX-300 with the expected availability factor greaterthan 95%. The BWRX-300 is capable to operate under the load following range of 50 to 100% with a ramp
rate of ±0.5% per minute.

7. Instrumentation and Control System
The BWRX-300 I&C system (also referred to as the Distributive Control and Information System or “DCIS”)
is a completely integrated control and monitoring system for the power plant. The overall I&C system isdivided into systems-Safety Class 1, Safety Class 2/3 and non-Safety.

93

8. Plant Layout Arrangement
The reference site for BWRX-300 is
a 260 m by 332 m footprint. The
power block is comprised of theReactor Building (RB), Turbine
Building (TB), Control Building(CB) and Radwaste Building (RwB).
The footprint of the power block isapproximately 140 m x 70 m.
The RB is the only Seismic Category
1 structure in the BWRX- 300. The
PCV and RPV are situated at or
below grade within the RB. A waterpool sites above the PCV. The three
Plant Layout Arrangement
ICS pools sit next to the pool above
the PCV with one IC located in each pool. The CB houses the control room, electrical, control andinstrumentation equipment. The TB encloses the turbine, generator, main condenser, condensate and feedwater
systems, condensate purification system, and off gas system.

9. Testing Conducted for Design Verification and Validation
For most structures, systems and components (SSCs), the BWRX-300 uses commercial off-the-shelfequipment and proven construction techniques. Design verification is performed to ensure the design complies
with customer requirements, technical requirements, regulatory requirements, and codes and standards.
Engineering tests that provide product performance or design basis information, or impact the design,
licensing, or operation of a product, product line, or environmental qualification, are also used. The test
program also includes prototype qualification tests, production tests, proof tests before installation, andpreoperational tests to demonstrate satisfactory performance of the system/components.

10. Design and Licensing Status
The BWRX-300 has undergone preapplication reviews with the United Kingdom Office of Nuclear Regulation(ONR), the United States Nuclear Regulatory Commission (U.S. NRC) and the Canadian Nuclear Safety
Commission (CNSC). In the United Kingdom, the BWRX-300 has undergone an evaluation by UK
Department for Business, Energy and Industrial Strategy funded Mature Technology. In the USA, five
Licensing Topical Reports (LTRs) for design features and analysis methods considered to have higher
regulatory risks have been submitted and approved. In Canada, the BWRX-300 is undergoing a combined
Phase 1 and 2 Vendor Design Review process that is expected to be completed in the Fall of 2022. In late2022, Ontario Power Generation (OPG) is planning to submit a License to Construct Application for a BWRX300 at Darlington New Nuclear Project site, DNNP-1.

11. Fuel Cycle Approach
The BWRX-300 has the same open fuel cycle as operating BWRs. Refueling outages are 10-20 days basedon the number of fuel assemblies to be replaced considering plant cycle duty. The cycle durations of 12 to 24
months can be accommodated to meet customer needs. Approximately 32 bundles are replaced following a
12-month cycle and 72 bundles following a 24-month cycle.

12. Waste Management and Disposal Plan
The BWRX-300 utilizes the lessons learned and best practices from the decades of operational experience of
the BWR fleet in waste management and disposal. Waste that is generated will be segregated for optimaltreatment, storage and final disposal. Gaseous wastes will be removed from the steam system by steam jet air
ejectors and passed through charcoal beds for absorption.

13. Development Milestones
2017
2018
2019
2020
2021
2022
2028

94

Conceptual design started
BEIS funded Mature Technology Evaluation the UK ONR
Pre-application activities initiated with the US NRC
Phase 1 and 2 Vendor Design Review initiated with CNSC in Canada
BWRX-300 Downselected by OPG after extensive evaluation of all SMRs
Planned OPG submittal of License to Construct for DNNP-1
Planned Commercial operation of lead BWRX-300

SMR-160 (Holtec International,
United States of America)
All content included in this section has been provided by, and is reproduced with permission of Holtec International, United States of America.

Parameter

Reactor Coolant System

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type

Holtec International, United Statesof America
PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

525 MW(t)174 gross / 160 net MW(e)Natural circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

15.5 / 3.4

Number of fuel assemblies in
the core
Fuel enrichment (%)

57

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

45 (initial design)

Reactivity control mechanism

Control rods and soluble boron

Approach to safety systems
Design life (years)

Fully Passive; utilize varied
phenomena and redundancy.
80

Plant footprint (m2)

28,000RPV height/diameter (m)15 / 3

RPV weight (metric ton)

295 (with fuel and internals)

Seismic Design (SSE)

0.5g, derived from the NRC
Regulatory Guide 1.60
Approximately 1/3 batch fraction
removed each refuelling outage
Passive safety cooling systems andactive non-safety systems; criticalcomponents below grade. Integrated
dry spent fuel storage andtransportation system
PSAR for commercial project by
2023, with detailed design tocomplete by 2025

Fuel cycle requirements /ApproachDistinguishing features
Containment Systems

Design status

243 / 321
UO2 pellet / square array

4.0 (average)

24 (equilibrium)

1. Introduction
The SMR-160 has been developed by Holtec International as an advanced PWR small modular reactor
producing 525 MW thermal power or 160 MW net electric power. The plant design incorporates robust passive
safety systems to achieve a highly reliable design that protects owner’s investment from all postulated
accidents, sabotage, or inadvertent human actions. In accord with Holtec’s Design Philosophy, the SMR-160
is designed as a ‘walk-away safe’ power plant – no operator actions are necessary to cope with design basis
accidents and safely reject decay heat. The plant is greatly simplified relative to conventional plants to improve
its fabricability, constructability, and maintainability, substantially facilitated by incorporating entirely passive
safety systems with a natural circulation primary loop NSSS. A modular construction plan for the SMR-160
involves fabrication and assembly of the largest shippable components prior to arrival at a site. A 24-month
construction period is envisaged for each nth-of-a-kind unit.

95

2. Target Application
The primary application of SMR-160 is electricity production with optional cogeneration equipment (i.e.,hydrogen generation, thermal energy storage, district heating, seawater desalination). The design is readily
configurable for siting in water-scarce locations using Holtec International’s proprietary air-cooled condenser
technology. The SMR-160 is capable of both “black-start” and isolated operation, rendering the plant ideal for
destinations with unstable power grids or off-grid applications.

3. Design Philosophy
Defence-in-Depth is achieved by including passive safety cooling systems and active non-safety systems
within the NSSS, with all critical components installed below grade and protected by a robust containment
enclosure structure. The plant is designed to be walk away safe, with a small spent fuel pool within containment
and a small source term, resulting in an orders of magnitude improvement in safety, e.g., CDF, as comparedto current generation plants, with no dependence on operator actions.

4. Main Design Features
Design Philosophy
The SMR-160 design philosophy is driven by the principal criterion of achieving unparalleled safety without
reliance on active systems or operator actions during design basis accidents, while ensuring the SMR-160
design remains inherently securable, fabricable, constructible, and economically competitive in world-wide
markets.

Nuclear Steam Supply System
The SMR-160 is a PWR with a naturally circulating reactor coolant system (RCS) primary loop. The RCS iscomprised of the reactor pressure vessel (RPV) and a steam generator (SG) in an offset configuration with an
integrated pressurizer welded to the top of the steam generator. The RPV and the SG are connected by a single
connection which contains both the hot leg and the cold leg in concentric ducts. The offset configuration allows
easy access to the core without moving the RPV or SG during refuelling. The entire primary circuit pressure
boundary is qualified under the ASME Code as a combined vessel, without any intermediary loop piping.

Reactor CoreThe SMR-160 employs an efficient reactor core design that uses a traditional reload shuffle. The reactor corecontains standard length 17x17 PWR fuel assemblies commercially available, along with typical magneticjack driven control rod assemblies. The reactor vessel internals support the reactor core, the control rod
assemblies, and the control rod drive shafts within the RPV. The core is designed for a nominal two-year cyclewith flexibility for shorter or longer cycles depending upon utility requirements.

Reactivity Control
Long term reactivity control is provided by burnable absorbers integral to the fuel which are designed to
optimize 3D power distributions, cold shutdown margin, and hot excess reactivity. Short term changes inreactivity are controlled by adjusting soluble boron and movements of control rod assemblies (CRAs). CRAs
are positioned by control rod drive mechanisms (CRDM) based on existing electro-mechanical technology.
The CRDMs are located outside the reactor coolant system on the RPV upper head.

Reactor Pressure Vessel and Internals
The RPV is an ASME Section III, Class 1, thick-walled cylindrical pressure vessel with an integrally weldedbottom head and a removable top head. The offset configuration of the SG and RPV enables the use of
traditional external control rod drive mechanism and greatly simplifies refuelling operations relative to typicalintegral PWR designs. The reactor internal structures are designed to be supported from the bottom of thevessel and are completely replaceable.

Reactor Coolant System
The SMR-160 RCS operates purely by natural circulation induced by the density difference in the primarywater and the height of the RPV and steam generator. There are no reactor coolant pumps in the system. The
RCS consists of three major components, a Reactor Pressure Vessel (RPV), a Steam Generator (SG), and an
integral Pressurizer, qualified as a single combine vessel.

Steam Generator
The SMR-160 includes a single, vertically oriented, once through straight tube SG with the reactor coolant
flowing inside thermally treated Inconel 690 tubes. The use of straight tubes ensures easy access for in-service
inspection. The SG uses sub-cooled feedwater to produce superheated steam on the shell side. The SG featuresa large inventory of secondary water on the shell side which provides substantial margin to dry-out.

Pressurizer
The pressurizer is integral to the steam generator and uses heaters and cold-water spray nozzles to perform the96

functions of a typical pressurizer. Integrating the pressurizer with the steam generator eliminates significant
primary piping. The large relative size of the Pressurizer eliminates any need for Power Operated Relief Valves(PORVs) and simplifies power manoeuvring.

5. Safety Features
The SMR-160 safety basis incorporates
defence-in-depth via multiple and varied
pathways for rejection of decay heat. All
safety systems are located inside therobust containment enclosure structure,rendering them secure and safe fromexternal threats. All makeup water needed
for a postulated loss of coolant accident(LOCA) is inside containment. Anotherlarge inventory of water within a captive
reservoir between the containment
enclosure structure and the containment
structure provides long-term postaccident coping and allows the decay heatremoval function to transition to air
cooling for an unlimited coping period
after a design basis accident without any
operator action or make-up.

PCCS and Power Conversion Diagram

Engineered Safety System Approach and Configuration
SMR-160 relies on passive and redundant safety systems that also operate by natural circulation. The passive
safety systems ensure safe shutdown can be maintained and decay heat removal occurs for an unlimited period
without the need for power, make-up water, or operator actions.

Passive Core Cooling System (PCCS) for Decay Heat Removal and Emergency Core Cooling
The PCCS is designed to provide emergency core cooling and makeup to the RCS during postulated accidents.
The system uses passive means such as natural circulation for core cooling and compressed gas expansion andgravity injection for core makeup without the use of active components such as pumps. The PCCS is comprisedof four major subsystems: (i) Primary decay heat removal system (PDHR); (ii) Secondary decay heat removal
system (SDHR); (iii) Automatic depressurization system (ADS) and (iv) Passive core make-up water system(PCMWS). The PDHR directly cools the primary coolant by re-routing the reactor coolant through a heatexchanger and rejecting the heat to a second loop full of water.

Spent Fuel Cooling Safety Approach
The spent fuel pool of the SMR-160 is located within the containment. The SMR-160 provides emergency
cooling to spent fuel during postulated events by integrating the cooling and make-up to the spent fuel pool to
a common reactor protection strategy. For long term cooling and make-up, the spent fuel pool acts as a
combined containment sump for the recirculation of coolant within the containment vessel to the reactor..

Containment and the Passive Containment Heat Removal System (PCHR)
The SMR-160 containment system consists of a steel containment structure (CS), enclosed within a reinforced
concrete containment enclosure structure (CES). The CES provides shielding and protection from external
events. In addition to preventing the release of radioactive fission products to the environment, the containment
system acts as a large passive heat exchanger. The PCHR passively cools the containment volume, without
any required actuations. The large heat transfer area and high conductance of the metal containment wall resultsin near-instantaneous heat rejection to the AR. The AR then rejects heat to the environment.

6. Plant Safety and Operational Performances
The SMR-160 natural convection-driven reactor coolant loop is coupled with an optimized simple steam cycle
and is well adapted to load following. The SMR-160 is designed to achieve a high capacity factor of > 95%with an estimated refuelling outage of 10 days every two-year fuel cycle. The reliance on passive safety
features and elimination of operator actions for response to postulated events yields a safer plant. The coredamage frequency is calculated to be on the order of 10-8/year and importantly on the order of 10-7/year when
considering only plant safety systems.

7. Instrumentation and Control Systems
SMR-160 utilizes the Mitsubishi Electric Total Advanced Controller (MELTAC) platform for the plant
I&C/HSI design. MELTAC provides unique nuclear specific I/O and configuration flexibility to perform all
nuclear safety and non-safety functions using the same digital platform..

97

8. Plant Layout Arrangement
Containment Enclosure Structure
The SMR-160 reactor is housed within acontainment structure (CS) protected by acontainment enclosure structure (CES)
comprised of steel-concrete modules. The
CES is missile-hardened and protects the CS
and safety systems from the most severe
environmental hazards or sabotage. Nearly
half of the CS and CES is embedded
underground. These structures house all
safety systems and the spent fuel pool.

SMR-160 Plant Layout

Reactor Auxiliary Building
The reactor auxiliary building houses many of the plant auxiliary systems. This building is designed to processspent fuel for dry interim on-site storage within Holtec International’s proprietary HI-STORM UMAX
modules (an underground dry cask storage technology), without any modification to the standard plant design.
The integrated HI-STORM UMAX canister system can also be used for off-site transport to a centralized
repository without repackaging.

Balance of Plant
The steam turbine and associated systems are housed within the turbine building structure at grade level. The
SMR-160 features a side exhaust steam turbine, optionally configured for air cooled condensation. The electric
power system consists of the main generator, main transformer, auxiliary transformers, non-safety diesel
generators and Class 1E batteries. The electrical system is designed to permit isolated operation in “islandmode” as well as start-up operations independent of the grid or “black-start.”

9. Testing Conducted for Design Verification and Validation
A thermal hydraulic test facility has been engineered to yield data and information for benchmarking the SMR160 following the process defined by the U.S. Nuclear Regulatory Commission. The facility provide data forsystem performance characteristics and important thermal hydraulic phenomena to support verification andvalidation. Facility implementation is underway in 2022, with experiments at an Integral and Separate Effects
Test (ISET) facility at Idaho National Lab (INL) in the U.S planned to begin in late 2023.

10. Design and Licensing Status
Pre-application activities for the SMR-160 have commenced with multiple international regulators in parallelwith development of commercial project opportunities. Formal pre-application activities, submittals andreviews are conducted with the USNRC in accord with a cooperatively agreed Regulatory Engagement Plan.

11. Fuel Cycle Approach
The SMR-160 fuel cycle is designed to discharge approximately one third of the fuel assemblies in the coreeach refuelling cycle, along with shuffling of a portion of the remaining fuel assemblies. The spent fuel isstored briefly in the spent fuel pool, which is uniquely protected within the same containment as the reactor.
New fuel assemblies are delivered using Holtec International’s HI-STORM system, which has decades ofoperating experience throughout the global light water reactor fleet. The HI-STORM system has received
multiple licensing approvals from the U.S. Nuclear Regulatory Commission.

12. Waste Management and Disposal Plan
High level waste management and disposal for the SMR-160 uniquely benefits from the integration of Holtec
International’s dry storage technologies. After removal of spent fuel from the spent fuel pool within a MultiPurpose Canister called an MPC-37, all spent fuel for the life of the plant can be stored on-site within an arrayof HI-STORM UMAX modules (an underground vertical storage cask design). The MPC-37 is a dual-purpose
canister licensed for transportation off-site within the HI-STAR 190 transportation overpack.

13. Development Milestones
2012
2015
2020
2023

98

Conceptual design of SMR-160 commencement
Conceptual design completed for SMR-160
Preliminary design completed for SMR-160
Ready for commercialization using a construction permit based process

Westinghouse SMR (Westinghouse ElectricCompany LLC, United States of America)
All content included in this section has been provided by, and is reproduced with permission of, Westinghouse Electric Company LLC, USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor typeWestinghouse ElectricCompany LLC, USA
Integral PWR

Coolant/moderator

Light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

800 / >225

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

15.5

Number of fuel assemblies in thecore
Fuel enrichment (%)

89

Forced circulation

294 / 324
UO2 pellet/17x17 square<5Core Discharge Burnup (GWd/ton)> 62

Refuelling Cycle (months)

24

Reactivity control mechanism

CRDM, boron

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

65 000RPV height/diameter (m)28 / 3.7

Seismic Design (SSE)

Based on CEUS sites

Distinguishing features

Incorporates passive safetysystems and proven componentsof the AP1000 plant and earlier
Westinghouse designs
Conceptual design completed

Design status

1. Introduction
The Westinghouse small modular reactor (SMR) is an integral pressurized water reactor (PWR) design thatbuilds upon the concepts of simplicity and advanced passive safety demonstrated in the AP1000® plant. The
power station delivers a thermal output of 800 MW(t) and a net electrical output of greater than 225 MW(e) asa standalone unit, completely self-contained within a compact plant site. The entire plant is designed formodular construction with all components shippable by rail, truck, or barge.

2. Target Application
The target application is the clean and safe generation of electricity; however, the Westinghouse SMR can also
be used to provide process heat, district heat, and off-grid applications, including the generation of power
necessary to produce liquid transportation fuel from oil sands, oil shale, and coal-to-liquid applications.

3. Main Design Features
Design of the Westinghouse SMR utilizes passive safety systems and proven components – realized in the
AP1000 plant reactor design and earlier Westinghouse designs – to achieve the highest level of safety,
resiliency, and certainty in licensing, construction, and operations. The Westinghouse SMR is designed to be
100 percent modular and limits the size of primary components in order to enable unrestricted transportation,
which reduces the need for costly infrastructure and increases the number of possible sites.

99

4. Main Design Features
Nuclear Steam Supply System
The Westinghouse SMR utilizes a light water pressurized reactor coolant system (RCS) that is integrated intoa single component, which eliminates the large-break loss-of-coolant accident (LOCA) from the postulated
event.

Reactor Core
The Westinghouse SMR reactor core is based on the licensed Westinghouse robust fuel assembly (RFA) design
and uses 89 standard 17×17 fuel assemblies with a 2.4 m active fuel height and Optimized ZIRLO® cladding
for corrosion resistance. A metallic radial reflector is used to achieve better neutron economy in the core while
reducing enrichment requirements to less than the existing statutory limit of 5.0 weight percent 235U.
Approximately 40% of the core is replaced every 2 years with the objective to achieve efficient and economical
operating cycle of 700 effective full power days, which coincides with existing regulatory surveillanceintervals.

Reactivity Control
Reactivity is controlled using the Westinghouse-developed system known as MSHIM™ control strategy ormechanical shim. MSHIM uses grey rods for short-term power control and boron dilution to adjust for fuelburnup over the longer term. Wireless instrumentation comprised of hardened electronics and reactor control
rod drive mechanisms (CRDMs) used in the Westinghouse SMR are based on proven AP1000 plant designs
but modified to allow for placement within the harsh environment of the reactor pressure vessel (RPV). This
proven design eliminates CRDM penetrations through the RPV head to prevent postulated rod ejection
accidents, as well as the potential for nozzle cracking, which has negatively impacted currently operating
plants. The upper internals of the RPV support 37 of these high-temperature-resistant, internal CRDMs forreactivity control during load-follow and similar operations.

Reactor Pressure Vessel and Internals
The RPV and reactor internals are designed to facilitate factory fabrication and shipment from the fabrication
facility. Designed to meet the requirements of the American Society of Mechanical Engineers (ASME) Boiler
and Pressure Vessel Code, these components are derived from existing Westinghouse products but redesigned
to function within the integral reactor assembly. The upper internals are an integral assembly containing all
the instrumentation and electrical penetrations to facilitate removal during refuelling.

Reactor Coolant System
The Westinghouse SMR design incorporates eight seal-less canned motor pumps, which are mounted
horizontally to the shell of the RPV just below the closure flange to provide forced reactor coolant flow throughthe core. A central primary riser directs the coolant flow as it exits the core to the steam generator. The reactor
vessel downcomer acts as the channel for delivering the coolant flow from the reactor coolant pumps to thecore inlet. The steam generator utilizes straight tubes with the primary reactor coolant passing through theinside of the tubes and the secondary coolant passing on the outside. An integral pressurizer is located above
the steam generator within the RPV to control pressure in the primary system. The moisture separation
functions typically performed in the steam generator occur in the SMR design in a separate steam drum locatedoutside of containment, reducing the reactor and containment vessel heights by approximately 6 meters. The
steam generator/pressurizer assembly can be removed for refuelling operations through a bolted closure flange
near the top of the integral reactor vessel.

5. Safety Features
The Westinghouse SMR is an advanced passive plant where the safety systems are designed to mitigate
accidents by natural driving forces such as gravity flow and natural circulation flow. The plant is not reliant
on alternating current (ac) power or other support systems to perform its safety functions. The 7-day minimum
coping time following loss of offsite power is a fundamental advancement over the 3-day coping time appliedin the operating plants. The integral reactor design eliminates large loop piping and potential large break LOCA
and reduces the potential flow area of postulated small-break LOCAs. The below grade locations of the reactor
vessel, containment vessel, and spent fuel pool provide protection against external threats and natural
phenomena hazards. The small size and low power density of the reactor limits the potential consequences ofan accident relative to a large plant. The plant is designed to be standalone, with no shared systems, thus
eliminating susceptibility to failures that cascade from one unit to another in a multi-unit station. The result isa plant capable of withstanding natural phenomena hazards and beyond-design-basis accident scenarios,
including long-term station blackout.

Engineered Safety System Approach and Configuration
The Westinghouse SMR is designed with passive safety systems that utilize the natural forces of evaporation,condensation, and gravity. The design basis and licensing of passive systems were first implemented in thedesign of the AP1000 plant. Elements of these systems are described in the following sections.
100Decay Heat Removal System
Three diverse decay heat removal methods are provided in the Westinghouse SMR. The first method of decay
heat removal uses gravity feed from the steam drum through the steam generator for approximately 80 minutesof natural circulation cooling. In this scenario, steam is released to the atmosphere through two redundant
power-operated relief valves. The second decay heat removal method can be achieved by cooling the RCS
with a passive decay heat removal heat exchanger, one of which is located in each of four core makeup tanks(CMTs). Heat from the CMTs is then rejected to four heat exchangers located in two ultimate heat sink (UHS)system tanks. The UHS tanks are sized to provide a minimum of 7 days of decay heat removal, with additional
options to replenish lost inventory and cool the plant indefinitely. A third diverse method of decay heat removal
capability is available by cooling the RCS with diverse bleed-and-feed methods, including a two-stage
automatic depressurization system that vents the RCS to the containment through direct vessel injection (DVI)
pathways, water injection from the four CMTs and in-containment pool (ICP) tank paths, and gravity-fed boric
acid tank water makeup to the DVI paths. The steam vented from the RCS to the containment is cooled andcondensed by the containment shell. The containment shell is cooled by the water in the outside containment
pool (OCP) that completely surrounds the containment. When the OCP water eventually boils, makeup water
is provided by gravity from each of the two redundant UHS tanks that maintain the OCP full of water. The
water condensed on the containment shell flows back into the RCS through two sump injection flow paths.

Below-Grade Location of Westinghouse SMR Reactor andContainment Vessels.

Three Diverse Decay Heat Removal Methods of theWestinghouse SMR.

Containment System
The containment vessel is a carbon steel vessel that is normally submerged in a pool of water. Pressure in thecontainment vessel following postulated events is maintained by transferring heat through the shell to the water
surrounding it. As this water boils, the inventory is made up from the two large UHS tanks that supply theplant with enough decay heat removal capacity for more than 7 days.

6. Plant Safety and Operational Performances
The design of the Westinghouse SMR represents a significant advancement in plant safety with an estimatedcore damage frequency of 5E-8 per reactor year while maintaining an expected capacity factor of 95%.

7. Instrumentation and Control Systems
An Ovation™-based digital instrumentation and control (I&C) system controls the normal operations of theplant. The protection and safety monitoring system provides detection of off-normal conditions and actuationof appropriate safety-related functions necessary to achieve and maintain the plant in a safe shutdown
condition. The plant control system controls non-safety-related components in the plant that are operated fromthe main control room or remote shutdown workstation. The diverse actuation system is a non-safety-related,diverse system that provides an alternate means of initiating reactor trip and actuating selected engineered
safety features.
Ovation is a trademark or registered trademark of Emerson Electric Company. Other names may be trademarksof their respective owners.

8. Plant Layout Arrangement
Reactor Building
The Westinghouse SMR main control room is also located completely below grade on the nuclear island;
additionally, there are multiple security monitoring stations located in separate sectors

Balance of Plant
The balance of plant design (BOP) consists of a conventional power train using a steam cycle and a watercooled (or optional air-cooled) condenser. The BOP operation is not credited for design basis accidents. The101

power conversion system is comprised of the turbine-generator, main steam system and condenser. Eachreactor drives a separate turbine generator and no sharing of reactor safety systems. The design includes “Island
Mode” capability to handle grid disconnection and is capable of 100% steam bypass capability to handle
turbine trip, both preventing a need for a reactor scram.

Plant Layout of Westinghouse SMR

Turbine Generator Building
The electrical generator is designed for air cooling, which eliminates the potential for explosions that can occur
with hydrogen-cooled options. The Westinghouse SMR condenser design includes the capability to use air
cooling. The turbine is designed to accommodate a wide variety of backpressures with different blade
configurations optimized for narrow-range, high-performance power. The water intake requirements will becomparable to existing plants on a per-power basis, but significantly less on a plant basis because of the lower
power rating. This low water usage enables the reactor to be sited in places previously not available for nuclear
construction.

Electric Power Systems
The Westinghouse SMR onsite power system consists of a main ac power system and a direct current (DC)power system. The main ac power system is a non-Class 1E system and does not perform any safety-related
functions. The plant dc power system is composed of the independent Class 1E and non-Class 1E DC powersystems. Safety-related dc power is provided to support reactor trip and engineered safeguards actuation.
Batteries are sized to provide the necessary dc power and uninterruptible ac power for items such as protection
and safety monitoring system actuation; control room functions, including habitability; DC-powered valves in
the passive safety-related systems; and containment isolation. Two diverse, non-safety AC power backup
systems are provided: 1) diesel-driven generators to provide power for defence-in-depth electrical loads, and2) a decay heat-driven generator. The decay heat-driven generator provides ac power to the plant using theheat generated by the core following reactor trip.

9. Design and Licensing Status
The Westinghouse SMR concept design has made substantial progress in support of U.S. and UK licensing.
Westinghouse is considering a number of business models for the successful deployment of the Westinghouse
SMR product globally. In addition, in February 2015, the U.S. Nuclear Regulatory Commission (NRC)
approved Westinghouse’s testing approach for the Westinghouse SMR design. The NRC approval is asignificant step toward design certification and will reduce the time ultimately needed to license theWestinghouse SMR. In a letter dated February 27, 2015, the NRC told Westinghouse that it had granted a
Safety Evaluation Report for the licensing topical report that the company submitted in April 2012 for agency
review and approval. The topical report, developed by a panel of experts from inside and outside ofWestinghouse, identified what would occur in the unlikely event of a small-break LOCA in the Westinghouse
SMR. It also defined the test program that Westinghouse will conduct in the future to prove that its safetysystems would safely shut down the reactor in response to a small-break LOCA. As a major technicalinnovation, the potential for intermediate and large-break LOCAs is eliminated in the Westinghouse SMR
design because there are no large penetrations of the reactor vessel or large loop piping.

10. Development Milestones
2015

102

Conceptual design completed

mPower (BWX Technologies, Inc.,
United States of America)
All content included in this section has been provided by, and is reproduced with permission of, BWX Technologies, Inc., USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

BWX Technologies, Inc.,
United States of America
Integral PWR

Coolant/moderator

Light water / Light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

575 / 195

NSSS Operating Pressure (primary),
MPaCore Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

14.8

Number of fuel assemblies in the core69

Forced circulation

290.5 / 318.9
UO2 pellet / 17x17 square

Fuel enrichment (%)<5

Core Discharge Burnup (GWd/ton)< 40

Refuelling Cycle (months)

24

Reactivity control mechanism

Control rods

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

157 000

RPV height/diameter (m)27.4 / 4.15

Seismic Design (SSE)

0.3g

Distinguishing features

In-vessel control rod drives
mechanism
Conceptual design

Design status

1. Introduction
The mPower™ plant consists of an integral PWR small modular reactor and related balance of plant, designedby Generation mPower LLC to generate a nominal output of 195 MW(e) per module. In its standard plantdesign, each mPower plant comprises a ‘twin-pack’ set, or two mPower reactor modules, generating a nominal
390 MW(e). The design adopts internal steam supply system components, once-through steam generators,pressurizer, in-vessel control rod drive mechanisms (CRDMs), and vertically mounted canned motor pumps
for its primary cooling circuit and passive safety systems. The plant is composed of reactor modules that are
fully shop-manufactured, rail-shippable to a site and installed into the facility.

2. Target Application
The primary application for the mPower reactor is electricity production. The mPower design could be
retrofitted to support other heat-requiring industries, desalination or co-generation applications.

3. Design Philosophy
The mPower design is based on the use of systems and components with an advanced plant architecture that
reduces licensing and construction risks. The mPower design employs passive safety features according to thedefence-in-depth principle, including an underground steel containment vessel structure and an undergroundspent fuel storage pool.

4. Main Design Features
Nuclear Steam Supply System
The NSSS consists of a reactor core, a steam generator (SG), reactor coolant pumps (RCPs), pressurizer andthe core internals that are integrated within the reactor pressure vessel (RPV). The NSSS forging diameterallows greater sourcing options and rail shipments.
103

Reactor Core
The reactor core consists of 69 fuel assemblies (FAs) that have less than 5% enrichment, Gd2O3 spiked fuelrods, Ag-In-Cd (AIC) control rods, and a design minimum 3% shutdown margin. The FAs are of aconventional 17×17 design with a fixed grid structural cage. FAs are shortened to an active length of 2.4 m
and optimized to maximize fuel utilization. The operational cycle is 24 months with a fuel burn cycle of up to
48 months.

Reactivity Control
Soluble boron is eliminated from the reactor coolant. The primary means of reactivity control for the mPower
design is achieved through the electro-mechanical actuation of control rods. The CRDM is fully submerged in
the primary coolant within the RPV boundary which precludes the possibility of control rod ejection accident
scenarios. Additional reactivity control is achieved by the strong negative moderator temperature coefficientby control of the secondary side feedwater flow rates.

Reactor Pressure Vessel and Internals
The mPower RPV houses the steam generator, CRDMs, pressurizer, reactor coolant pumps and the isolation
valves. The integrated RPV inherently eliminates the possibility of a large break loss-of-coolant accident(LOCA). Reactor internals include core support, internal structures and all structural and mechanical elements
inside the RPV.

Reactor Coolant System
The primary cooling mechanism of the mPower reactor under normal operating condition and shutdown
condition is by forced circulation of coolant. The reactor uses eight reactor coolant pumps (RCPs) located on
a 360° pump shelf at the top of the coolant riser. The large reactor coolant system (RCS) volume of the mPower
reactor allows more time for safety systems to respond in the event of an accident. Additional cooling water ispassively provided via the Emergency Core Cooling System (ECCS) for continuous cooling to protect the coreduring a small break LOCA.

Steam Generator
The steam generator (SG) is located within the annular space formed by the inner RPV walls and the riser
surrounding and extending upward from the core. The upper vessel assembly including the SG is removed foraccess to the core during refuelling and allows for inspection and maintenance in parallel with fuel exchange.

Pressurizer
The integrated electrically heated pressurizer located at the top of the RPV maintains a nominal 14.8 MPa.
Reactor coolant pressure is controlled by the heaters and steam space spray.

5. Safety Features
The integral reactor is contained within a steel containment vessel located fully underground within the Reactor
Service Building to provide enhanced protection against external events. The mPower plant safety features
meet a seven-day coping time without off-site power.

Engineered Safety System Approach and Configuration
The inherent safety features of thereactor design include a low corelinear heat rate that reduces fuel andcladding
temperaturesduringaccidents, a large RCS volume thatallows more time for safety system
responses in the event of an accident,
and small penetrations at highelevations, increasing the amount of
coolant available to mitigate a small
break LOCA. The mPower plant
deploys an enhanced spent fuel poolconfiguration, which is installed
underground, with a large heat sink tocope up for 30 days in case of loss offuel pool cooling.

Decay Heat Removal System
The mPower reactor deploys a decay heat removal strategy, with an auxiliary steam condenser on thesecondary system, water injection or cavity flooding using the refuelling water storage tank, and passive
containment cooling.

104

Emergency Core Cooling System
The mPower ECCS is a safety system that provides three basic functions: (1) depressurization of the RCS, (2)reactor coolant inventory control during the event, and (3) core decay heat removal. With a system of automatic
depressurization valves (ADVs) and the large coolant reserve provided by accumulators referred to asintermediate pressure injection tanks (IPIT) and the in-containment refuelling water storage tank (RWST), thereactor core remains covered following a design-basis event. The IPITs are maintained with pressurized
nitrogen over the water. Should the automatic depressurization valves open, the reactor pressure vessel will
depressurize until in equilibrium with the containment atmosphere. During and following that pressure
equalization, check valves between the reactor pressure vessel and IPITs (early) and the RWST (late) open,
injecting gravity-driven cooling water from the RWST. The RWST provides a minimum of 7 days to as muchas 14 days of cooling without the need for external intervention or AC power to maintain reactor core coolingand safe shutdown.
To address the single-failure general design criteria, the ECCS is designed for n+1 components, with allcomponents located inside containment. There are four high-pressure and four low pressure depressurization
paths arranged in pairs connected to four pressurizer connections. The ADVs are the isolation valves for theeight lines. There are two injection paths to the RCS. Each path to the RCS is connected to an IPIT and to the
RWST.

Containment System
The steel containment vessel and interfacing safety systems work in concert to protect the core, provide longterm core cooling, and prevent the release of radioactive materials to the environment without reliance on AC
power or operator action well beyond the regulatory expectation of 72 hours following an accident. This
seismic Category I structure is designed to withstand the maximum internal pressure from design basis
accidents, including LOCA and steam line break. Normal access to the steel containment vessel is via
personnel hatches and a removable equipment hatch provides access for large component replacement. Internal
containment atmosphere pressure and temperature is maintained by passive containment cooling (PCC) via an
integral water tank situated in direct contact with the containment dome, providing passive cooling under
accident conditions. Heat is removed from the hot steam and air inside containment via heat transfer throughthe containment dome structure to the water in the PCC tank located on the outside surface of the dome.
Containment atmosphere response to a breach in the RCS may be characterized by two distinct phases. The
first phase (blowdown) is an injection of hot steam corresponding to RCS depressurization. The high rate ofsteam injection during this period increases containment atmospheric temperature and pressure and quickly
disperses a steam and air mixture throughout the containment volume. With cold walls and other structuresurfaces, condensation is the most important heat transfer mechanism. Natural convection and conductionlimited heat transfer to the PCC tank distinguishes the second phase. There is sufficient water available topassively remove a minimum of seven days of core decay heat by evaporation.

6. Plant Safety and Operational Performances
Moderating and maximizing the time response of event loads relative to their limits is a focal point in
improving the reactor inventory and cooling safety functions. The total inventory and its distribution
throughout the system factor into this assessment. Further, reserve primary coolant from interfacing safetysystems, most notably the RWST, can extend these time response periods both temporally and to a broader
range of off-normal plant states. The arrangement of reactor core and steam generator thermal centers is crucial
to the plant's capability to remove heat by natural circulation following a loss of forced circulation. By
vertically separating these two components within an integral pressure vessel, the design of the mPower reactor
encourages this natural convection heat removal rather than requiring engineered pump-driven systems. The
mPower design includes analogous circuits to remove heat from the secondary and from the containment
systems. In the former instance, the system is a non-safety, defence-in-depth system, providing the capacity
for long-term decay heat removal. In addition, the reactor coolant inventory and purification system (RCI)
serve as an active, non-safety decay heat removal system. System response derived from LOCA simulations
has demonstrated that core temperatures are well below limits.

7. Instrumentation and Control Systems
The instrumentation and control (I&C) system provides the capability to monitor, control and operate plantsystems. It functions to (1) control the normal operation of the facility, (2) ensure critical systems operate
within their designed and licensed limits, and (3) provide information and alarms in the control room for theoperators. Important operating parameters are monitored and recorded, during both normal operations andemergency conditions to enable necessary operator actions. The I&C system is implemented using modern,scalable digital technology. Protection functions are implemented in a dedicated layer, which actuatesengineered safety features if required to ensure safety of the facility. A second layer provides automatic control
functionality and includes the capability for operators to control all systems within the plant. A third layer
provides monitoring of all plant parameters, with advanced data processing capability to enable efficient
operations. This layered architecture provides a high degree of automation while ensuring safety.

105

8. Plant Layout Arrangement
Reactor Building
The reactor service building is a reinforced concrete, seismic Category I structure that surrounds the steel
containment vessel and spent fuel storage pool which are located below grade level. The control room islocated below grade in the reactor service building and contains the control system and operator interface forboth reactors.

Balance of Plant
The balance of plant design (BOP) consists of a conventional power train using a steam cycle and a watercooled (or optional air-cooled) condenser. The BOP operation is not credited for design basis accidents. The
steam and power conversion system is comprised of the turbine-generator, main steam system and condenser.
Each reactor drives a separate turbine generator and no sharing of reactor safety systems. The design includes“Island Mode” capability to handle grid disconnection and is capable of 100% steam bypass capability tohandle turbine trip, both preventing a need for a reactor scram.
i. Turbine Generator Building:
The turbine generators are housed in a separate building. The water-cooled condenser provides for a nominal
output power of 195 MW(e). The turbine-generator is designed for power manoeuvring and flexible gridinterface. Turbine-generator support systems include a turbine bearing lubrication oil system, an electrohydraulic control system, a turbine gland seal system, turning gear, over speed protective devices, a generator
rectifier section and a voltage regulator.
ii. Electric Power Systems:
The main generator supplies power to plant auxiliaries during normal plant operation through an isolated phasebus duct and the unit auxiliary transformer. Offsite power to plant auxiliaries during startup, shutdown, andoutage conditions is supplied via back-feed from the main transformer and the unit auxiliary transformer, withthe generator circuit breaker open. If the unit auxiliary transformer is not available, offsite power is supplied
via the station service transformer.

9. Design and Licensing Status
BWX Technologies, Inc. (formerly, The Babcock & Wilcox Company) and Bechtel Power Corporation are
members in a formal alliance called Generation mPower LLC organized to design, license and deploy mPower
modular plants. In 2013, the mPower program became the first recipient of funding under the US Departmentof Energy SMR Licensing Technical Support public-private cost-share program. Design engineering activities
in support of a Design Certification Application continue at BWXT and Bechtel to further develop thetechnology with a focus on design certification. Design Certification and site-specific licensing is expected tobe completed in order to support an initial deployment in the mid-2020s.

10. Development Milestones
2009
2010
2012
2014
2015

2016

106

BWX Technologies, Inc. (formerly B&W) officially introduced the mPower SMR
concept
Pre-application design certification activities engagement with the United States
Nuclear Regulatory Commission
The Integrated System Test (IST) facility located in Bedford County, Virginia, was put
into operation
Tennessee Valley Authority (TVA) announced its intention to submit an Early Site
Permit Application at the TVA Clinch River site in Roane County for two or more
SMR modules.
The Babcock & Wilcox Company spun off its Power Generation business and theremaining company changed its name to BWX Technologies, Inc., retaining its interestin Generation mPower LLC and related nuclear steam supply system (NSSS) design
authority
BWXT and Bechtel Power Corporation agree to a Framework Agreement which
provides for transition to a new management structure with Bechtel responsible for
Program Management of the mPower programOPEN20 (Last Energy Inc., United Statesof America)
All content included in this section has been provided by, and is reproduced with permission of, Last Energy Inc., United States of America.

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer,
country of origin

Last Energy Inc., United States of America

Reactor type1-loop PWR

Coolant/moderator

Light water / Light water

Thermal/electrical capacity,
MW(t)/MW(e)73 / 22

Primary circulation

Forced circulation

NSSS operating pressure(primary/secondary), MPa

15.5 / 2.8

Core inlet/outlet coolant
temperature (oC)

286 / 312

Fuel type/assembly array

UO2 pellet / 17x17 array

Number of fuel assemblies
in the core

Up to 24

Fuel enrichment (%)<4.95

Refuelling cycle (months)

72

Core discharge burnup(GWd/ton)>30

Reactivity control
mechanismApproach to safety systems

Control rods

Design life (years)

40+ years

Plant footprint (m2)2023

RPV height/diameter (m)9.1 / 2.3

RPV weight (metric ton)

89.3

Seismic design (SSE)

0.5g horizontal and 0.4g vertical peakground accelerations

Fuel cycle
requirements/approach

Open cycle; wet and dry storage on-site.
Provisions for closed cycle are underconsideration.
Fully-modular construction, undergroundnuclear island, digital controls, closed
tertiary loop using air coolingDistinguishing features

Design status

Passive, Defense-in-depth, Diverse, andRedundant

Detailed design

1. Introduction
Last Energy has developed a production ready, fully-modular micro-PWR with nominal electric output of
20MWe for distributed baseload applications including industrial siting and combined heat and power. Last
Energy's technology is a scaled-down one-loop pressurized water reactor (PWR) that specifically avoids
innovations around fuel, reactor physics, or material science. Despite the small power output, the fuel is still
standard PWR fuel, <5% enriched, found in over 300 currently operating PWRs, and the entire power plantfollows a design requirement of incorporating only "off-the-shelf" equipment. The primary technicalinnovations are specific to productization, modularization, and constructability of the power plant. The values
listed in this book are generally for OPEN20, an academic version of Last Energy’s PWR-20 commercial plantdesign.

107

2. Target Application
The plant is designed for both electricity production as well as non-electrical process heat applications. Each
plant comes standard with combined heat, power and cooling capabilities, and can deliver up to 300C processsteam for broad industry applications.

3. Design Philosophy
The plant design has been developed to target four key outcomes:(1) Minimize technology risks by utilizing existing and proven technology (e.g., PWR & steam)(2) Straightforward demonstration of risk reduction using conservative assumptions and fewer critical systems(3) Modular design and factory construction accelerates delivery time by reducing required on-site activities(4) Minimize siting constraints and environmental impact (e.g., air-cooled, small footprint)

4. Main Design Features
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) is based on the PWR technology found in over 300 currently
operating reactors. The NSSS includes the reactor system, reactor coolant pumps, a pressurizer, a steam
generator, misc. equipment, and piping required to generate steam for the turbine. The system uses forced
convection and operates in a single closed primary loop.

Reactor Core
The nuclear fuel is industry standard UO2 enriched up to 4.95%, clad with a zirconium alloy, and arranged ina 17x17 assembly. The core contains up to 24 fuel assemblies delivering a thermal power of approximately 70
MW(t). Each fuel assembly contains poisoned fuel pins using distributed Gd2O3 as well as 24 guide tube
locations for control rod fingers and a central instrument tube. The operational characteristics are designed toensure large margin to thermal-mechanical fuel limits including DNBR. The core is designed for a nominal
six-year cycle with flexibility for shorter cycles depending upon utility requirements.

Reactivity Control
Core reactivity is controlled by means of control rods, solid burnable poison, and soluble boron dispersed in
the primary coolant. Burnable poison rods flatten the radial and axial power profile, which results in an
increased thermal margin of the core. The number and concentration of the burnable absorber rods in each fueltype are selected so that reactivity of each assembly can be as flat as possible. The control rods use boron
carbide as the absorber material.

Reactor Pressure Vessel and Internals
The base of the reactor pressure vessel (RPV) is forged carbon steel. It is a thick-walled cylindrical pressure
vessel with an integrally welded bottom head and a removable top head. The RPV has a relatively small width
to height ratio. The internals are the same as a conventional PWR scaled to fit a reduced fuel bundle count.
Notably this is not an integral reactor system.

Reactor Coolant System
The Reactor Coolant System (RCS) is of the same basic design of light water PWRs in use throughout theworld today and includes the equipment and piping required to transfer energy produced from nuclear fuelassemblies to convert high pressure water into steam. The system utilizes a single closed loop forced circulation
to transfer high pressure (approximately 15.5 MPa), high temperature (approximately 330C) water between
the reactor vessel and the steam generator equipment where it passes through tubes in shell and tube heatexchangers to transfer energy and generate steam (approximately 315C) from lower pressure feedwater.

Steam Generator
A vertically oriented shell and tube steam generator uses heat created by the reactor to produce steam for theturbine. This style of heat exchanger isolates the reactor coolant loop (flow through the tubes) from the turbine
steam (flow through the shell).

Pressurizer
The reactor coolant pressure is established and kept nearly constant by a pressurizer. Electric heaters create
and control the size of the steam space inside the pressurizer vessel. This imposes backpressure on the reactor
coolant to prevent it from boiling at other locations. The pressurizer is sized to greatly dampen the effects oftransient pressure events.
108Primary pumps
After exiting the steam generator, coolant is pumped to the reactor by vertical, canned, single stage centrifugal
pumps that operate with variable frequency drives. To improve system reliability there are three reactor coolant
pumps; however, only one pump is required to deliver full coolant flow to the reactor for normal operations.

5. Safety Features
Engineered Safety System Approach and Configuration
Last Energy's design has been developed to incorporate the best practices learned over five decades of nuclear
plant engineering. Defense-in-depth is provided through multiple layers of fault prevention (e.g., IAEA 5 levelstandards) as well as independent and diverse active and passive systems. Passive safety systems are designed
to perform without human input or electrical power.

Decay Heat Removal System
A passive decay heat removal system is included to protect the system integrity during unlikely accident
scenarios.

Emergency Core Cooling System
Last Energy's design has been developed to incorporate the best practices learned over five decades of nuclear
plant engineering. Defense-in-depth is provided through multiple layers of fault prevention (e.g., IAEA 5 levelstandards) as well as independent and diverse active and passive systems. Passive safety systems are designed
to perform without human input or electrical power.

Spent Fuel Cooling Safety Approach / System
The plant design includes provisions for interim on-site wet and dry spent fuel storage in preparation for final
offsite disposal or reprocessing.

Containment System
The nuclear island is contained inside a modular, leak tight, steel enclosure that prevents any inadvertent
release of radionuclides during either normal operating or accident conditions. Additionally, the subterranean
containment provides protection from external events.

6. Plant Safety and Operational Performances
The behavior of the plant during normal and faulted conditions has been analyzed and assessed using industry
validated codes to demonstrate significant safety margins across the levels of defense-in-depth. The
Probabilistic Safety Assessment calculates a LERF < 1E-7. Additionally, our capacity factor is greater than
95% and refueling outage time is less than 3 months every 72 months.

7. Instrumentation and Control System
A distributed control system collects measurements and issues commands via input/output devices in theprocess areas. Digitized data is communicated over a secure Ethernet-based network. The on-site control roomhas only interfaces for critical process information and alarms in keeping with the largely autonomous
operational concept of the plant. Extensive plant data from multiple installations is transmitted via one-way
communication links to a remote monitoring center where off-site staff can observe trends and instruct field
personnel to investigate abnormalities or perform corrective actions. Remote control of plant equipment is notpossible.

8. Plant Layout Arrangement
Two primary structures are located on-site.
The nuclear island is located below grade andhouses the NSSS. The balance of plantstructure is located above grade and houses all
remaining systems including the steam turbine
generator, feedwater system, auxiliary
systems, and electrical systems.

9. Testing Conducted for Design
Verification and Validation
Last Energy's power plant does not include any

Artistic rendering of OPEN20 containment design109experimental systems. Its NSSS is similar in configuration to several historical micro/modular, land-based,
single-loop PWRs such as the BR-3 and MHA-1 reactors. Where it deviates from traditional nuclear systems(e.g., digital controls) it utilizes only equipment that can be found in thousands of comparable fossil-based
thermal power plants.

10. Design and Licensing Status
The PWR-20 has completed detailed design and is producing construction ready drawings. Pre-licensing
activities have commenced with multiple international nuclear regulators in parallel alongside the fabricationof non-nuclear systems. The project execution plan projects initial operation of the first deployed reactors by
the mid-2020s. Site control has already been secured at multiple locations.

11. Fuel Cycle Approach
The reactor is operated on a 72-month fuel cycle. The plant design includes provisions for interim on-site wet
and dry spent fuel storage in preparation for final off-site disposal or reprocessing.

12. Waste Management and Disposal Plan
All spent fuel is stored onsite until final decommissioning. Spent fuel assemblies and waste can be handled
and transported in manner similar to that of conventional PWRs.

13. Development Milestones
2018
2020
2021
2022
2025
2026
2027

110

Utility and government "need-finding" through Titans of Nuclear podcast
Launch of OPEN100 (open source PWR) for technical and economic validation
Preliminary studies and pre-conceptual design for a 20MWe scale reference plant
Detail design complete and pre-licensing overviews with four (4) nuclear regulators
Commercial operation of reference plant
10x 20MWe plants online
Assembly line manufacturing begins

PART I.2. WATER COOLEDSMALL MODULAR REACTORS(MARINE BASED)

111

112

KLT-40S (JSC “Afrikantov OKBM”, RussianFederation)
All content included in this section has been provided by, and is reproduced with permission of JSC “Afrikantov OKBM”, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

JSC “Afrikantov OKBM”,
Rosatom, Russian Federation
PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

150 / 35NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

12.7

Number of fuel assemblies in
the core
Fuel enrichment (%)

121

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

45.4

Reactivity control mechanism
Approach to safety systems

Control rod driving
mechanism
Active (partially passive)

Design life (years)

40

Plant footprint (m2)4320 (Floating NPP)RPV height/diameter (m)4.8 / 2.0

RPV weight (metric ton)

N/A

Seismic Design (SSE)

9 point on the MSK scale

Distinguishing features

Floating power unit forcogeneration of heat andelectricity; no onsite
refuelling; spent fuel takebackConnected to the grid in Pevek
in December 2019. Enteredfull commercial operation in
May 2020.

Design status

Forced circulation

280 / 316
UO2 pellet in silumin matrix

18.6

30-36

1. Introduction
The KLT-40S is a PWR developed for a floating nuclear power plant (FNPP) to provide capacity of 35 MW(e)per module. The design is based on third generation KLT-40 marine propulsion plant and is an advanced
version of the reactor providing the long-term operation of nuclear icebreakers under more severe conditions
as compared to stationary nuclear power plant (NPP). The FNPP with a KLT-40S reactor can be manufactured
in shipyards and delivered to the sites fully assembled, tested and ready for operation. There is no need todevelop transportation links, power transmission lines or the preparatory infrastructure required for land based
NPPs, and there is a high degree of freedom in selecting the location for a FNPP as it can be moored in any
coastal region.

2. Target Application
The FNPP with KLT-40S is intended to provide cogeneration capabilities for power and heat supply to isolated
consumers in remote areas without centralized power supply. Besides, this FNPP can be used for seawater
desalination as well as for autonomous power supply for sea oil-production platforms.

113

3. Design Philosophy
KLT-40S is the reactor for Akademik Lomonosov FNPP, intended for reliable power and heat supply to isolated
consumers in remote areas without centralized power supply and where expensive delivered fossil fuel is used.

4. Main Design Features
Nuclear Steam Supply System
The steam lines while exiting from the SGs are routed through containment to a set of steam inlet valves, andfinally into the turbine building for electricity conversion. Cogeneration equipment could be modified into themedium-low temperature heat process concept if one or multiple separation heat exchangers are positioned
between the primary and secondary loops.

Reactor Core
Fuel utilization efficiency is achieved by using dispersion fuel elements. One of the advantages foreseen by
the FNPP under construction is long term independent operation in remote regions with decentralized power
supply. The design requires refuelling of reactor after every 2.5–3 years of operation. Refuelling is performed
14 days after reactor shutdown when the levels of residual heat releases from spent FAs have reached therequired level. The spent nuclear fuel is initially stored on board at the FNPP and then returned to Russianfederation. No special maintenance or refuelling ships are necessary. Single fuel loading is done in order toprovide maximum operation period between refuelling. The fuel is loaded in the core all at once with all fuelassemblies being replaced at the same time.

Reactivity Control
The control rod drive mechanism (CRDM) is electrically driven and releases control and emergency control
rods into the core in case of station black out (SBO). The speed of safety rods driven by electric motor, in thecase of emergency is 2 mm/s. The average speed of safety rods being driven by gravity is 30 – 130 mm/s.

Reactor Pressure Vessel and Internals
The KLT-40S reactor has a four-loop forced and natural circulation coolant loop; the latter is used only in theemergency heat removal mode. This reactor is utilized at all operating nuclear icebreakers.

Reactor Coolant System
The reactor has a modular design with the core, steam generators (SGs) and main circulation pumps connectedwith short nozzles. The reactor has a four-loop system with forced and natural circulation, a pressurizedprimary circuit with canned motor pumps and leak tight bellow type valves, a once-through coiled SG andpassive safety systems. KLT-40S thermal-hydraulic connections comprising external pressurizer,
accumulators, and separation heat ex-changer are in proximity of the reactor systems. The pressurizer is not
an integral part of the reactor systems and in this design it is formed by one or more separate tanks, designed
to accommodate changes in coolant volume, especially severe during reactor start-up. The core is cooled bycoolant flowing from core bottom to top, in accordance with typical PWR core flow patterns. However, flow
patterns between the core shroud and the RPV inner walls differ significantly from conventional external loop
PWR configurations. Once hot coolant exits the top of the core and enters any of the multiple SGs, it uses
coaxial hydraulic paths wherein the cold and hot legs are essentially surrounding one another. As hot coolant
enters the SG, it begins to transfer thermal energy with the fluid circulating in the secondary loop (secondaryside of the SGs).

5. Safety Features
The KLT-40S is designed with proven safety aspects such as a compact structure of the SG unit with shortnozzles connecting the main equipment, primary circuit pipelines with smaller diameter, and with provenreactor emergency shutdown actuators based on different operation principles, emergency heat removal
systems connected to the primary and secondary circuits. Additional barriers are provided to prevent the releaseof radioactivity from the FNPP caused by severe accidents. Among them are passive and active physicallyseparated and independent safety systems, I&C systems, diagnostic systems, active cooling train throughprimary circuit purification system’s heat exchanger thermally coupled with a ’third’ independent circuitexchanging heat energy with ambient sea or lake water, active cooling train through the SGs heat exchangers
with decay heat removal accomplished through the condenser which in turn is cooled down by ambient sea orlake water, 2 passive cooling trains through the SGs with decay heat removal via emergency water tank heatexchangers, and venting to atmosphere by evaporation from said tanks. Both active and passive safety systems
are to perform the reactor emergency shutdown, emergency heat removal from the primary circuit, emergency
core cooling and radioactive products confinement. The KLT-40S safety concept encompasses accident
prevention and mitigation system, a physical barriers system, and a system of technical and organizational
measures on protection of the barriers and retaining their effectiveness, in conjunction with measures onprotection of the personnel, population and environment. The KLT-40S safety systems installed on FNPPs are
distinctive from those applied to land-based installations in security of the water areas surrounding the FNPP,
anti-flooding features, anti-collision protection, etc. Passive cooling channels with water tanks and in-built
114heat exchangers ensure reliable cooling to 24 hours.

Engineered Safety System Approach and Configuration
The active components of the protection system are scram actuators for six (6) groups of the control rods.

Decay Heat Removal System
The decay heat removal system is intended to remove core residual heat upon actuation of reactor emergency
protection in case of abnormal operation including accidents, as well as to remove residual heat at normal RP
decommissioning. The decay heat removal system includes two secondary passive cooling channels via steam
generators, one active secondary cooling channel via steam generators and one active cooling channel via theprimary/third heat exchanger.

Emergency Core Cooling System
The ECCS is intended to supply water to the reactor for core cooling in accidents associated with primarycoolant loss, makeup of primary coolant during process operations, supply of liquid coolant to the reactor at
failure of the electromechanical reactor shutdown system, adjustment of water chemistry and hydraulic testingof the primary circuit and associated systems, secondary and third loop sections disconnected at inter-circuit
leaks and designed for primary pressure. The ECCS includes high-pressure ECCS subsystem with makeup,
high-pressure ECCS subsystem with hydraulic accumulators and Low-pressure ECCS subsystem withrecirculation pumps.
Passive system of containment emergencypressure decrease (condensate system)

Passive emergency core cooling system(hydraulic accumulators)

Active emergency corecooling system

Passiveemergency
shutdown
cooling system

From STP (Standby and emergency
feed water pumps)

Active system ofliquidabsorber

Active emergency
shutdown coolingsystem (throughprocess condensers)
Containment passive
emergency pressure
decrease system (bubbling)

Active emergency corecooling system
Recirculation system pumps
System of reactor caisson
filling with waterContainment System
The containment for the KLT-40S is configured for FNPP applications and is made of steel shell designed tosustain mild pressurization, while the reactor systems are positioned inside a reinforced ‘reactor room’ whose
bottom forms a steel-lined tank. This tank can be flooded with cooling water for decay heat removal as well
as for shielding purposes. The top portions of the reactor room can be pressurized as the reactor room is plugged
by a steel and concrete plug. Once removed, the plug provides access to the reactor systems and to the core forrefuelling or maintenance operations.

6. Plant Safety and Operational Performances
The KLT-40S NPP ensures electricity and heat generation within the power range of 10% to 100% for acontinuous operation of 26 000 hours. The NPP is designed for manoeuvring speed of up to 0.1 %/s. As acountermeasure against the external impact, the NPP is fitted with both ground safety and floating physicalprotection means. Structures are designed to be placed in the Arctic zone at the depth of 2 m at freezing
temperatures. The FPU and NPP buildings are designed to withstand the crash of an aircraft of 10 tons. Basedon analysis, the radiation emission limits are satisfied for all conditions.

7. Electric Power Systems
The electric power system in the FPU is comprised of the following: main electric system; and emergency
electric system. The main electric system of the FPU is intended to generate electricity and transmit it to thepower system of the region, as well as to transmit electricity to internal consumers. The system includes two
main three-phase AC generators of 35 MW each and eight back-up diesel generators of 992 kW each. The
emergency electric system supplies electricity to safety system loads in all operation modes, including loss ofoperating and back-up electric power sources. The FPU has independent emergency electric systems for eachreactor plant. Each emergency electric system has two channels with an emergency diesel generator of 200
kW.
115

8. Plant Layout Arrangement
The coastline line of the FCNPP has the complex engineering building with equipment to distribute andtransmit electricity to loads and to prepare and transfer heating water to loads and auxiliary buildings,
including: two hot water storage tanks; partially in ground tank with slime water; wet storage bunker; twocooling towers; access control point; site enclosure; lighting towers. The coastal line of the FCNPP does notprovide for handling nuclear materials and radiation hazardous media.

Reactor Building
The FPU is a flush deck non-self-propelled rack-mounted vessel with hull and multi-layer deckhouse. The
medium portion of the FPU has a reactor compartment and nuclear fuel handling compartment. A
turbogenerator compartment and electrotechnical compartment are arranged in the ship’s head with respect tothe reactor compartment, auxiliary installations compartment and accommodations are arranged in astern. Eachreactor plant is arranged within steel pressure containment, which is a reinforced structure of the FPU casing.
The containment is designed for maximum pressure, which can develop during accidents. Onboard the FPU,
storages for spent cores and means are arranged that ensure reactor reloading.

Control Building
The KLT-40S reactor is controlled using the operator’s automated workstation through respective control panel
located in the central control room. In case it is impossible to carry out control from the central control room,
information on the reactor status is obtained and safety systems are activated to make reactors subcritical andcontrol reactor plant cooling using emergency cooling control panels located outside the central control room.

General cross-section
view of the FPU

Turbine Generator Building
The steam turbine plant (STP) is intended to convert the thermal power from steam obtained in the KLT-40S
reactor to the electric and thermal one to heat water in the intermediate circuit of the cogeneration heating
system. The FNPP structure includes two steam turbine plants. Each STP is independent of the other and isconnected to its own module of KLT-40S. Heat is delivered to the shore by heating intermediate circuit water,which circulates between FPU and the shore, using steam from adjustable turbine steam extraction.

9. Design and Licensing Status
KLT-40S is the closest to commercialization of all available FNPP designs, and expects deployment throughthe Akademik Lomonosov FNPP. The KLT-40S is a modified version of the commercial KLT-40 propulsion
plants employed by the Russian icebreaker fleet. The environmental impact assessment for KLT-40S reactor
systems was approved by the Russian Federation Ministry of Natural Resources in 2002. In 2003, the first
floating plant using the KLT-40S reactor system received the nuclear site and construction licenses from
Rostechnadzor. The keel of the FNPP carrying the KLT-40S, the Akademik Lomonosov in the Chukotka
region, was laid in 2007. The construction of Akademik Lomonosov was completed in 2017. The Akademik
Lomonosov has started commercial operation in December 2019 in the town of Pevek in Chukotka region.

10. Development Milestones
1998
2002
2006
2012
2017
2018
Summer 2019
December 2019
May 2020

116

The first project to build a floating nuclear power plant was establishedThe environmental impact assessment was approved by the Russian Federation Ministryof Natural Resources
After several delays the project was revived by Minatom (Russian Federation Ministryof Nuclear Energy)
Pevek was selected as the site for the installation of NPPs. JSC “Baltiysky Zavod”
undertook charge of construction, installation, testing and commissioning the first FPU
Completion of construction and testing of the floating power unit at the Baltic shipyard
Dock-side trials, fuelling, final tests completion with reactor core, attainment of reactor’s
first criticality
Transportation of FPU to the town of Pevek
Connected to grid on 19th of December in Pevek
Fully commissioned in Pevek on 22nd of May

ACPR50S (CGNPC, China)
All content included in this section has been provided by, and is reproduced with permission of CGNPC, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of
China General Nuclear PoweroriginGroup (CGNPC), China
Reactor type
Loop type PWR
Coolant/moderator
Light water / light water
Thermal/electrical capacity,
200 / 50
MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15.5

Core inlet/outlet coolant
temperature (oC)299.3 / 321.8

Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup (GWd/ton)Reactivity control mechanism

UO2 pellet / 17 × 17 square
37

Approach to safety systems
Design life (years)
RPV height/diameter (m)Fuel cycle requirements/approachDistinguishing features
Design status

<5
30
< 52
Control rod driving mechanism(CRDM), solid burnable poison
Passive40
7.2 / 2.2
60 months
Floating power boat, once-throughsteam generator, passive safetysystemDetailed design

1. Introduction
The ACPR50S is a small modular offshore floating reactor developed by the China General Nuclear Power
Corporation (CGNPC) aiming for high safety and adaptability, modularized design, and multi-purpose
applications. It is intended as a flexible solution for combined supply of heat, electricity and fresh water formarine resource developments, energy supply and emergency support on islands and along the coastal area.

2. Target Application
As an offshore floating SMR, the ACPR50S is designed for the following multipurpose applications:combined energy supply for offshore oil drilling platform; offshore combined energy supply; coastland andisland combined energy supply; energy supply for offshore mining, nuclear power ship; and distributed clean
energy for islands together with solar energy and wind power.

3. Design Philosophy
The ACPR50S adopts design simplification to reduce cost and investment risks to be competitive withconventional offshore energy sources. Modular design is adopted through standardized and streamlined
manufacturing, aiming for shorter construction period as well as lower cost. A long refuelling cycle allows
for higher load factors.

4. Main Design Features(a) Nuclear Steam Supply System
The compact loop-type PWR nuclear steam supply system (NSSS) design of the ACPR50S consists of thereactor pressure vessel (RPV) that houses the core, two once-through steam generators (OTSG), two main
reactor coolant pumps (RCP) and a pressurizer (PZR), all of which are interconnected by short reactor
coolant system legs. The primary cooling system is based on forced circulation during normal operation. The
system has natural circulation capability and heat removal capacity up to 10% thermal power.

117

(b) Reactor Core
The low power density design with a low enriched UO2 fuelled core ensures a thermal margin of greater than
15% which can accommodate any anticipated transient event. This feature ensures the core thermal
reliability under normal and accident conditions. The 37 fuel assemblies (FAs) of ACPR50S core, with anaxial length of 2.2m, have a square 17 × 17 configuration. The expected average fuel enrichment is less than
5%, similar to standard PWR fuel. The reactor will be able to operate 30 months per fuel cycle.

(c) Reactivity Control
Core reactivity is controlled by means of control rods, solid burnable poison and soluble boron dispersed in
the primary coolant. Burnable poison rods flatten the radial and axial power profile, which results in an
increased thermal margin of the core. The number and concentration of the burnable absorber rods in eachfuel type are selected so that reactivity of each assembly can be as flat as possible. There are 16 control rods,
with a magnetic force type control rod driving mechanism (CRDM).

(d) Reactor Pressure Vessel and Internals
ACPR50S reactor pressure vessel is 2.2m in diameter and 7.2m high. It envelopes and fixes the core and the
RPV internals, so that the fission reaction of the nuclear fuel is limited in one space.

(e) Reactor Coolant System
The ACPR50S primary cooling under normal operating condition and shutdown condition is done by forced
circulation. The RCS has been designed to ensure adequate cooling of reactor core under all operational
states, and during and following all postulated off normal conditions. The two RCPs are connected to the
OTSG through short annular pipes, as are the two OTSGs to the RPV, therefore eliminated large bore piping
and reduced opening of the main equipment. The integral design of RCS significantly reduces the flow areaof postulated small break LOCA.

(f) Steam Generator
The ACPR50S has two
OTSGs with helicallycoiled tubes to produce
superheated steam undernormal operating conditions. The OTSGs arelocated on both sides of
the reactor vessel. The
small inventory of thesecondary side (tube side)
water in each OTSG
prohibits a return to power
following a steam line
Energy Cascade Supply
break. In the case of
accidents, the OTSG can be used as the heat exchanger for active and passive secondary residual heatremoval system (ASHR & PSHR), which remove the decay heat from the primary system.

(g) Pressurizer
The pressurizer of ACPR50S is located outside of the reactor vessel connected to one of the hot leg
connecting the RPV with a steam generator. The pressurizer is designed to control the system pressure atnearly constant level for normal plant operation due to the large pressurizer steam volume and the heater
control. As the volume of the pressurizer is designed sufficiently large, condensing spray is not required forthe load manoeuvring operation. The reactor over-pressure at the postulated design basis accidents relatedwith a control failure can be reduced through the actuation of the pressurizer safety valve (PSV).

(h) Primary pumps
The two ACPR50S Reactor Coolant Pumps (RCP) are intended to circulate the reactor coolant through thereactor core. They are mounted directly to Reactor Pressure Vessel by short nozzles. The reactor coolant
pumps are designed with no seals, eliminating the potential risks of seal failure LOCA, which significantly
enhances safety and reduces pump maintenance.

5. Safety Features(a) Engineered Safety System Approach and Configuration
The ACPR50S is designed with passive safety systems that comprise passive safety injection system (SIS),
automatic depressurization system (ADS), Passive Secondary Residual Heat Removal System (SHR),
containment pressure suppression system (CPS), Passive Containment Heat Removal System (CHR),
containment and containment isolation system (CIS), containment hydrogen control and filtration exhaust118system (CHE).These passive safety systems are used to cope with design basis accidents (DBA) and severe
accidents with core melts.

(b) Decay Heat Removal System
The SHR is developed to remove the decay heat if the normal decay heat removal pathway is unavailable
under accident condition. For non LOCA events, SHR removes the decay heat of the core through natural
circulation between the OTSG and the SHR heat exchanger. This decay heat is ultimately removed by thecooling water in the cooling water tank outside the containment eventually. If power supply is still available
and, if the normal decay heat removal pathway fails under non LOCA accident, the feed water system will bestarted instead of the SHR.

(c) Emergency Core Cooling System
The SIS is a very important engineered safety system which is developed to cool and boride the corefollowing DBA and extends the no operator action time of plant operators to 7 days. Its main function is tocontrol and mitigate the consequences of the accident and prevent the DBA from becoming a serious beyond
design basis accident (BDBA). The core cooling is completely driven by the natural force following the DBA
such as LOCA, which simplifies the system's composition and operation greatly. Both high-pressure andlow-pressure safety injection are driven by gravity, and the medium pressure safety injection is driven bycompressed gas. In order to effectively connect the high-pressure, medium-pressure and low-pressure safety
injection, ADS is used.

(d) Containment System
The containment system is to contain radioactive material and protect the environment against primarycoolant leakage. The containment system of ACPR50S is called reactor cabin, a square containment. The
containment has a volume of 870 m3, a design internal pressure of 1.4 MPa and a design external pressure of
0.3 MPa. The containment isolation system (CIS) is to provide isolation for the containment and to preventand restrict the escape of radioactive fission products in the event of an accident. The CHE is used to reducethe concentration of hydrogen in the containment to the safety limit under DBA and BDBA and tocontinuously monitor the hydrogen concentration at the top of the containment. The CPS is developed tocope with the DBA that can lead to a pressure rise inside containment and suppress the containment pressure
peak to ensure the integrity of the containment. The CPS is composed of the suppression pool system andsuppression pool cleaning and cooling system. The CHR is used to prevent the containment slow overtemperature and overpressure. The CHR is a passive natural circulation system and can provided 7 days ofcooling for the containment in the case of no external water apply.

6. Plant Safety and Operational Performances
ACPR50S is suitable for all kinds of sea states such as swing, concussion, vibration, typhoon, seaquake andso on. The operational power range can be justified from 20% to 100% nuclear power (Pn) and can be
operated steadily for a long period for any power level of 20% to 100% Pn to satisfy the power demand.

7. Instrumentation and Control System
The I&C system design for ACPR50S is based on defence in depth concept, compliance with the single
failure criterion and diversity. The I&C system design for ACPR50S is mainly used in the steady-state andtransient power during operation and provides automatic protection against unsafe reactors and abnormal
operation, and provide the trigger signal to mitigate the consequences of accident conditions. Two reactors
share one control room, one technical support centre, and two separate remote shutdown stations to ensure
control and operation of the plant under normal and accident conditions.

8. Plant Layout Arrangement
A single reactor module with the electrical, the steam generator, and auxiliary nuclear facilities is installedinside a non-propelled barge for sharing facilities and reduced cost. Plant main building consists of thereactor containment cabin, the nuclear auxiliary cabin, the emergency diesel generator cabin and the turbinegenerator cabin. For efficient radiation management, the plant main building is sub-divided into two zones,
the duty zone and the clean zone. Systems linked with refuelling, overhauling, radwaste treatment are
installed in the onshore basement.
(a) Floating Platform
The reactor building and fuelstorage area are equipped with afull monitoring system with closed
circuit monitoring system tooversee and prevent unauthorizedaccess to the fuel. Reactor
building is a pre-stressed concrete
shell structure composed of a right

General Plant Arrangement in the float platform

119

cylinder with a hemispherical dome, with steel plate lining to act as a leak tight membrane. Reactor building
is founded on a common base-mat together with the auxiliary building in which the main control room andfuel storage area are located. The figure above shows the overview of the reactor cabin, auxiliary cabin,turbine-generator cabin, the emergency diesel generator cabin and the main control room arrangement of
ACPR50S.
(b) Onshore Basement
The onshore basement of ACPR50S houses the fuelling building, the radwaste treatment building, and otherbalance of plant (BOP) buildings. Refuelling and overhauling is performed in the onshore basement.
(c) Control Room
The compact control room is designed for one man operation under normal conditions of the plant and islocated in the ship (offshore). The main control room (MCR) is a key facility to cope with any emergency
situations, so it is designed to ensure that plant personnel successfully perform the tasks according to theproper operating procedures. To achieve these goals, human factors engineering (HFE) process andprinciples are applied and verified using the full scope dynamic simulator.

9. Testing Conducted for Design Verification and Validation
The CGN’s experiment centre has safety test platforms, equipment and key technology test platform, SMR
wave condition test platform, etc. Various experiments are performed for design validation.

10. Design and Licensing Status
The ACPR50S has completed the preliminary design work and is currently carrying out detailed design. An
industrial demonstration plant of ACPR50S is planned to be constructed in China.

11. Fuel Cycle Approach
The self-propelled experimental reactor adopts the whole heap refuelling strategy. All 37 fuel assemblies in
the full core are discharged after a fuel cycle, then 37 new fuel assemblies are loaded. The core loading
scheme adopts high leakage loading, that is, the components with high enrichment degree are arranged in theouter ring of the core, and the components with low enrichment degree are arranged in the inner ring of thecore to flatten the radial power distribution. The spent fuel assemblies are placed in the spent fuel pools.

12. Waste Management and Disposal Plan
Liquid radwaste system is designed to prevent or minimize the creation of radioactive liquid effluent, andthis achieved, wherever possible, by internal recycling. Gaseous radwaste system is designed to minimize theradioactivity associated with the resulting environment discharge. Solid radwaste system is designed tocollect, preliminarily treat and temporarily store the solid wastes during the operation. And the wastes will besent to disposal site for final disposal.

13. Development Milestones
2012
2014
2020
2022

120

Starting Conceptual Design and formulating plan of theoretical tests
Completion of Overall Design
Completion of Preliminary Design
Detailed design stageACP100S (CNNC/NPIC, China)
All content included in this section has been provided by, and is reproduced with permission of, CNNC/NPIC, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
CNNC/NPIC, China
country of origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Integral PWR
Light water / light water
385 / 125

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
15 / 4.6

Core inlet/outlet coolant
temperature (oC)

286.5 / 319.5

Fuel type/assembly array

UO2 / 17×17 square pitch arrangement
arrangement
57

Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)
Reactivity control mechanism

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approachDistinguishing features

Design status< 4.9524< 52
Control rod drive mechanism(CRDM), Gd2O3 solid burnable poisonand soluble boron acid
Passive + Active
40
8 000 (Floating NPP)
10 / 3.35
300
Not applicable
Shore refuelling, temporally stored in
the spent fuel poolIntegrated reactor with once throughsteam generator, towed to sites along
coast
Basic design

1. Introduction
ACP100S was developed based on its land-based reactor ACP100 whose pilot project has been underconstruction in Hainan province in China to provide capacity of 125 MW(e) per module. It inherits most of
the main features of ACP100 with integrated design and could be used for power generation, desalination,
heating, for offshore and open sea area, and isolated island, etc. It mainly adopts passive safety system withactive passive measure as an implement to cope with the marine environment. It could be constructed in the
shipyard and delivered to the sites, and satisfy the safety requirement of third generation PWR.

2. Target Application
The ACP100S is a multipurpose power reactor designed for coastal and open sea areas, or isolated island where
central electric grid is hard to reach. It could be used for electricity production, heating, steam production orseawater desalination.

3. Design Philosophy
The ACP100S realizes design simplification by integrating the primary cooling system and enhanced safety
by using the cooling water from the sea. It adopts mature design philosophy with land-based ACP100 being
constructed. ACP100S could be constructed in the shipyard and factory, which greatly improve the economic
competitiveness.
1214. Main Design Features
Nuclear Steam Supply System
The integrated nuclear steam supply system (NSSS) design consists of the reactor core, and sixteen (16) oncethrough steam generators (OTSG). The four (4) canned motor pumps are installed nozzle to nozzle to the RPV.
All these technologies provide high inherent safety and avoid large LOCA in large-size nuclear power plant.

Reactor Core
The 57 fuel assemblies (FAs) of ACP100S core with total length of 2.15 m core have a squared 17×17
configuration. The fuel 235U enrichment is about 1.9 – 4.95%. The reactor will be able to operate 24 months atbalance fuel cycle.

Reactivity Control
The reactivity is controlled by means of control rods, solid burnable poison and soluble boron dissolved in the
primary coolant. There are 20 control rods, with a magnetic force type control rod driving mechanism (CRDM).

Reactor Pressure Vessel and Internals
The RPV and equipment layout are designed to enable the natural circulation between reactor core and steam
generators. The RPV is protected by safety relief valves against over-pressurization in the case of strong
difference between core power and the heat removed from the RPV.
The internals not only support and fasten the core but also form the flow path of coolant inside RPV.

Reactor Coolant System
The ACP100S primary cooling mechanism under normal operating condition and shutdown condition is doneby forced circulation. The RCS has been designed to ensure adequate cooling of reactor core under all
operational states, during and following all postulated off normal conditions. The integral design of RCS
significantly reduces the flow area of postulated small break LOCA.

Steam Generator
There are 16 OTSGs, which are mounted within the RPV. All the 16 OTSGs are figured in the annulus between
the reactor vessel and hold-down barrel. The bottoms of OTSGs are limited their position by the hole on barrel
supporting hub, the heads are welded to the reactor vessel steam cavity.

Pressurizer
The pressurizer of ACP100S is located outside of the reactor vessel. The pressurizer is a vertical, cylindrical
vessel with hemispherical top and bottom heads.

Primary pumps
The ACP100S uses canned-motor pumps as reactor coolant pumps which are directly mounted on the RPV
nozzle. The shaft of the impeller and rotor of the canned-motor pump is contained in the pressure boundary,eliminating the seal LOCA of reactor coolant pump.

5. Safety Features
Engineered Safety System Approach and Configuration
The ACP100S is designed with inherent features, eliminating large primary coolant piping which in turns
eliminates large break LOCA. It mainly adopts passive safety system with active passive measure as asupplement to cope with the marine environment. The safety system mainly consists of the passive decay heatremoval system (PDHRS), passive emergency core cooling system (ECCS), passive containment suppression
system (PCS) and active containment spray system.

Decay Heat Removal System
The PDHRS prevents core meltdown in the case of design basis accident (DBA) and beyond DBA, such as
station black out, complete loss of feedwater, small-break LOCA (i.e., to prevent the change of beyond DBA
to severe phase). The PDHRS of ACP100S consists of two residual remove exchangers and associated valves,piping, and instrumentation. The residual heat remove exchanger is located in the passive residual heat remove
water storage tank mounted on left and right side of shipboard, which provides the heat sink for the emergency
cooler. The decay heat is removed from the core by natural circulation. The PDHRS provides core cooling for
72 hours without operator intervention.

Emergency Core Cooling System
The emergency core cooling system (ECCS) consists of two safety injection system and recirculation system,
which mainly include two core make-up tanks (CMT), two low head pressure injection pumps and associatedinjection lines and valves. In the DBA, cooling water from the CMTs would be injected to the reactor core by
gravity force to cool down the reactor core, and recirculation would be generated for long term after the DBA
122

accident.

Spent Fuel Cooling Safety Approach / System
ACP100S adopts the mature shore refuelling strategy, and would be towed to the dock when refuelling isneeded after one refuelling period of 24 month.

Containment System
The ACP100S use small steel containment containing one module of the NSSS. It's a vertical, cylindrical
vessel with hemispherical top and bottom heads, a personnel gate is mounted on the cylindrical side and ahemishperical top gate is set as a refuelling gate. The containment is figured in the reactor compartment which
is a reinforced structure of the boat. The containment could withstand the maximum pressure during DBA of
NSSS, and also withstand the external water pressure in the sink accident.

6. Plant Safety and Operational Performances
Nuclear safety is always the first priority. The ultimate goal of nuclear safety is to establish and maintain an
effective defence that can effectively protect people, community, and environment from radioactive disaster.
To be specific, the design and operation of ACP100S ensures that radiation dose to the workers and to themembers of the public do not exceed the dose limits and kept it as low as reasonably achievable. Accident
prevention measures ensure that radioactive consequences are lower than limited dose in terms of all theconsidered accident sequences and even in the unlikely severe accidents, mitigation of accidents induced
influences can be ensured by implementing emergency plan. The design of ACP100S incorporates operational
experience of the state-of-the-art design. Proven technology and equipment are adopted as much as reasonably
possible. It's estimated that ACP100S has the core damage frequency (CDF) of less than 1×10-5 per reactor
year and the large release frequency (LRF) of 1×10-6 per reactor year. The load factor is no less than 0.9.

7. Instrumentation and Control System
The Instrumentation and Control (I&C) system designed for ACP100S is based on defence in depth concept,
compliance with the single failure criterion and diversity. The diversity in the design of I&C system is achieved
through: (1) different hardware and software platforms for 1E and N1E I&C, (2) reactor protection system(RPS) with functional diversity, and (3) diverse protection systems to cope with the common mode failure of
the RPS. I&C systems of the NSSS include reactor nuclear instrumentation system, RPS, diverse actuation
system, reactor control system, rod control and rod position monitoring system, reactor in–core instrumentation
system, loose parts and vibration monitoring system and other process control systems.

8. Plant Layout Arrangement
The ACP100S is a non-propelled FNPP mounted on the boat, and could be towed to locations where neededwith the ability of operating independently. Actually, one or two NSSS modules could be installed on the boat
according to the requirement, with the length of the boat about 200meters or 150 meters respectively. The
containment figured in the reactor compartment is located in the middle of the boat to lower the effect causedby the ocean. The turbogenerator compartment is located next to the reactor compartment, andaccommodations are arranged in astern. The ACP100S has enough space to figure reactor auxiliarycompartment and I&C room, the cogeneration facility could be installed on the board or at the coastline
according to the operation requirement.

9. Testing Conducted for Design Verification and Validation
Since ACP100S is developed based on mature ACP100，most of verification tests have been finished, suchas CHF, test relevant with SG, flow induced vibration of internals, etc. Experiments are mainly planned toverify the new design brought by the ocean environment. For example, the control rod drive line cold and hot
test, containment scale structure test, and containment suppression test are in progress. ACP100S has a mature
NSSS with only small part of verification needed.

10. Design and Licensing Status
Presently, the licensing of ACP100S is yet to be launched. The preliminary feasibility study of ACP100S has
been finished aiming at the site in city of Yantai in the East of China.

11. Fuel Cycle Approach
Spent fuel processing is conducted on the shore, using the service facility prepared at the base harbour, which
is similar to the land-based reactor.

12. Waste Management and Disposal Plan
After waste is transported to the base harbour, the waste management approach and disposal plan is similar toother nuclear power plants.

12313. Development Milestones
2012
2018
2019
2021
2023
2024

124

The concept design of ACP100S finishedThe basic design of ACP100S finished
Preliminary feasibility study for one site finished
Preliminary design of ACP100S finished
PSAR planned to submit to National Nuclear Safety Authority
Target to start constructionBANDI-60 (KEPCO E&C, Republic ofKorea)All content included in this section has been provided by, and is reproduced with permission of KEPCO E&C, Republic of Korea.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of
KEPCO E&C, Republic oforiginKorea
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)PWR
H2O (Light Water)
200 / 60

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced15.5 / 6

Core inlet/outlet coolant
temperature (oC)

290 / 325

Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup (GWd/ton)Reactivity control mechanism
Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements/approachDistinguishing features
Design status

Oxide / 17×17
52
4.95
48~60
29.4
Rods
Passive safety system
60
6 500
11.2 / 2.8110
N/A
4~5 years / Single batch 4.95%
Passive Safety System
Conceptual design

1. Introduction
BANDI-60 is a compact two-loop PWR with two Utube type steam generators. KEPCO E&C focuses on
deploying a floating nuclear power plant in the sea.
BANDI-60 is designed for niche markets, for instance,
distributed power and heat supply to remotecommunities, desalination, and hybrid energy systems
with renewables, through a collaboration with marine
and shipbuilding industries.

2. Target Application

Typical Plant Layout

BANDI-60 is designed for off-gid and remote areas such as islands, Arctic area and isolated grid.

3. Design Philosophy
The BANDI-60 design adopts technology innovations developed in-house from 2013, such as the in-vessel
control element drive mechanism (IV-CEDM), soluble boron-free (SBF) design and operation, and the TopMounted In-Core Instrumentations (TM-ICI). The CEDM inside of reactor pressure vessel will eliminate
control element ejection accident. BANDI-60 is designed to operate at soluble boron-free conditions tosimplify the design, operation and maintenance of nuclear reactor.

125

4. Main Design Features
Nuclear Steam Supply System
BANDI-60 is a block-type PWR plant in which the main components are
directly connected, nozzle-to-nozzle, thus eliminating the large break loss of
coolant accident (LBLOCA) from the design basis accident. This provides also
improved operational surveillance and maintenance as compared to integral
type designs. For steam generator, the U-tube recirculation type is adopted
considering its performance has been proven with plenty of operational
experience in commercial nuclear power plants.

Reactor Core
The BANDI-60 reactor uses a conventional UO2 ceramic fuel enriched up to
4.95 % in a 17×17 square array fuel assembly. To produce 200MW thermal
power over a 4 – 5 years, the reactor core is composed of 52 fuel assemblies.
BANDI-60 does not use soluble boron. The Pyrex (B2O3) rods as burnable
absorber are loaded in the core to control the excess reactivity over core life
by replacing 24 fuel rods with different boron concentrations. The numbers
shown in the figure represents boron concentration. The centre fuel assemblies
have a higher boron concentration up to 40% while as low as 5% for the outer
fuel assemblies to flatten radial power distribution. The burnup calculation
showed that the core can operate for 57.5 months with full power.

Reactor Cross-section View

Reactivity Control
The BANDI-60 reactor core uses 40 control rod assemblies with 24 B4C
control rods per assembly. In conjunction with the burnable poison, the Reactor Core with Burnable Poison
control rod assemblies can provide enough negative reactivity to maintain• S2 R4 •shutdown with margin under worst condition for the soluble boron free
S2
S1
R2
S1
R3
S2
BANDI-60 reactor core. There are two (2) control rod groups, 18 for regulatinggroup and 22 for shutdown group. The control group is further grouped in 4 • R3 • S2 R1 S2 S1 •sbanks, R1 through R4, with 4 control rod assemblies per each bank except R4
which has 6 assemblies. The shutdown group consists of 2 banks, 10 assemblies R4 S1 R1 R4 S1 • R2 S2for S1 and 12 assemblies for S2. The remaining 12 fuel assemblies without S2 R2 • S1 R4 R1 S1 R4control rod are reserved for the TM-ICI.

Reactor Pressure Vessel and Internals

•s S1

S2

R1

S2

•

R3

S2

R3

S1

R2

S1

S2

•

BANDI-60 Reactor Pressure Vessel (RPV) is similar to a conventional PWR• R4 S2 •vessel. The RPV is a vertically mounted cylindrical vessel with a hemisphericallower head welded to the vessel and a removable hemispherical Reactor Vessel
Control Rod Arrangement
Closure Head (RVCH). The RPV contains internal structures, core support
structures, fuel assemblies, the IV-CEDMs, and control and instrumentation components. The inlet nozzle of
RPV is directly connected to the outlet nozzle of Steam Generator (SG). The Direct Vessel Injection (DVI)
nozzles are attached to the RPV for a direct emergency coolant injection as a part of the safety injection system.
The lifetime of the RPV is extended to 60 years by the use of low carbon steel. The reactor internals consist of
the core support structures, which include the core support barrel, upper guide structure assembly, IV-CEDM
support structure, lower support structure, and the internal structures.

Reactor Coolant System
The primary coolant flows in through the RPV inlet nozzles from the reactor coolant pump, passes through theannulus between RPV and core support barrel, through RPV bottom plenum and core, and finally flows out
through the outlet nozzles of RPV connected to the inlet nozzle of SG.

Steam Generator
The steam generator, a U-tube recirculation type, is employed as the basic design since its performance has
been proven with plenty of operational experience in commercial nuclear power plants. new technologies will
be examined for their feasibility such as compact plate-shell steam generator.

Pressurizer
The pressurizer is integrated into the upper head of the reactor pressure vessel where a relatively large waterand steam volume is provided as compared to that of conventional nuclear power plants. The pressurizerpressure is controlled by heaters and sprays. The CEDM is installed inside the reactor vessel.

Primary pumps
The Reactor Coolant Pump (RCP) is equipped with a leak-tight canned motor.

126

5. Safety Features
Engineered Safety System Approach andConfiguration
The BANDI-60 passive safety system consists of the PSIS,PRHRS, and PCCS. The PSIS performs gravity driven safety
injection and system depressurization. The PRHRS performs decay
heat removal by means of natural circulation and steam
condensation. The PCCS performs condensation and convection
through the containment vessel to the ultimate heat sink andrefueling water tank (URWT) that has enough inventory of water
for more than a month.

Decay Heat Removal System

Safety Systems

The PRHRS provides decay heat removal when the normal cooling of RCS through SGs and condenser is notavailable after reactor trip. The steam from the SG condenses in the PRHRS heat exchanger, which is locatedinside the URWT, and the condensate returns by gravity to the SG via the feedwater nozzle. It cools down the
RCS to the safe shutdown condition and maintains it for an extended time without refilling the URWT. Theenergy released to the containment atmosphere is continuously removed to the water stored in the URWT and
CVH Cooling Jacket. This passive containment cooling will decrease the core damage frequency (CDF).

Emergency Core Cooling System
The PSIS provides an RCS inventory makeup for any loss of coolant events other than LBLOCA by a gravity
driven flow. The Core Makeup Tank (CMT) is pressurized to the RCS pressure by the Pressure Balance Line(PBL). Isolation valves between the CMT and DVI nozzle remain closed during normal operation, and areopened when the RCS depressurizes, the pressurizer level decreases, or the containment pressure increases
beyond their set points. When the RCS pressure further decreases, the coolant stored in the Emergency Core
Cooling Tank (ECCT) is injected by gravity. To facilitate ECCT injection into RCS, the RCS pressure has tobe reduced and balanced with the containment pressure. For this rapid depressurization of RCS, the Safety
Depressurization Valves (SDV) on the pressurizer are automatically opened before ECCT injection. Spilledwater from the break is collected at the bottom of the containment and eventually the reactor vessel becomes
submerged in the water and is continuously cooled by convection and natural circulation.

Spent Fuel Cooling Safety Approach / System
The storage and handling facilities for new and spent fuel are located in the right side auxiliary building.
Seismic category I tanks and pumps, equipment for the component cooling water system as well as equipmentof plant cooling water system are arranged in the left side auxiliary building.

Containment System
The containment vessel of BANDI-60 is a steel cylindrical type with 11 m inner diameter and 18.5 m in height.
The reactor building, which surrounds the containment vessel, incorporates features to minimize and mitigate
postulated severe accident phenomena. The steel containment vessel improves construction schedule compared
to conventional pre-stressed concrete design. The vessel is designed to enable modularity to further reducecomplexity and construction time. Heat released from the RCS to the containment is continuously removed to
the water stored in the URWT through metal wall during a postulated accident.

6. Plant Safety and Operational Performances
BANDI-60 is under conceptual design with CDF target of 10-8/RY and 10-9/RY of LERF.

7. Instrumentation and Control System
The instrumentation and control system (I&C) is composed of a fully digitalized system with data
communication network and adopts a simple and compact structure suitable for BANDI-60 as floating SMR
plant. This system can prevent unwanted reactor trip due to equipment failure or external disturbance by meansof a Reactor Trip Prevention System (RTPS). The I&C design prevents the possibility of software common
cause failure by implementing its own diversity design in the safety system. Therefore, a diverse actuation
system is not required resulting in a more compact I&C system.

127

8. Plant Layout Arrangement
All safety systems are located in the reactor building positioned at the center of the floating site (ship).
Auxiliary buildings are located on the left and right hand sides of the reactor building.

Sectional View of Plant Layout

Plan View of Plant Layouta) Reactor Building

The reactor building contains two containment vessels with reactor coolant system, part of engineered safetysystems, and related auxiliary systems. The reactor building is located in the middle of the ship, and the left
and right sides of the reactor building are surrounded by auxiliary buildings and the north and south sides areequipped with passages for equipment transfer or operator movement.

b) Auxiliary Building

The auxiliary building houses facilities necessary to perform the main functions of the reactor coolant systemand the engineered safety features related to the safety shutdown of the plant.

c) Turbine Building
Turbine building is adjacent to the left side auxiliary building and includes turbine, condenser, electricalsystem and plant chilled water system. It is designed to secure a space for carrying in and out of equipment,
connection to external electric line and onshore facilities connected the secondary system.

d) Access Control

The access control area is in the access control and living area adjacent to the right side auxiliary building witha space for land facility connection. It also contains access controls to facilities and radiation areas.

9. Testing Conducted for Design Verification and Validation
To be developed.

10. Design and Licensing Status
To be launched. Design and Licensing status will be launched and site permit for FOAK plant will be
developed. Design status of the BANDI-60 is under conceptual design.

11. Fuel Cycle Approach
The initial core design is based on a single batch operation having 4 – 5 years of cycle length to minimize
refueling outage. As a floating power plant, the new and spent fuel are transported from and to an inland fuelfabrication and storage facility using a dedicated fuel transport ship. The spent fuel is shipped to the inlandstorage facility after about 10 years (2 cycle lengths) in the on-board spent fuel storage. To optimize the fuelcycle cost, a multi-batch core can be adopted with a shorter cycle length.

12. Waste Management and Disposal Plan
As a floating power plant, the spent fuels are transported from and to an inland fuel fabrication and storage
facility using a dedicated fuel transport ship. The back-end fuel cycle option for spent fuel is dependent on theplant owner’s policy and requirements.

13. Development Milestones
2012
2013
2016
2018
2019
2024
2025
2030

128

Preliminary and basic studies
R&D projects for technology innovation started
NSSS conceptual design started
R&D projects for technology innovation completed
NSSS conceptual design completed BOP and floating system conceptual design
Complete BOP and floating system conceptual design
Complete basic design and licensing review
Projected construction

ABV-6E (JSC “Afrikantov OKBM”,
Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of JSC “Afrikantov OKBM”, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

JSC “Afrikantov OKBM”, Rosatom,
Russian Federation
PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

38 / 6 – 9

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

16.2

Number of fuel assemblies in
the core
Fuel enrichment (%)

121

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

N/A

Reactivity control mechanism

Control rod driving mechanism

Approach to safety systems

Passive

Design life (years)

40

Plant footprint (m2)

20 000 (basic design)

RPV height/diameter (m)6 / 2.4

Natural circulation

250 / 325
UO2 pellet/hexagonal< 20

120 – 144

Seismic Design

7 per Richter scale (basic design)

Distinguishing features

Natural circulation in the primarycircuitFinal design

Design status

1. Introduction
The ABV-6E is reactor plant (RP) as a part of nuclear power system (NPS) that produces 14 MW(t) and 6
MW(e) in cogeneration mode or 9 MW(e) in condensation mode. ABV-6E integral PWR adopts natural
circulation of the primary coolant. The ABV-6E design was developed using the operating experience of PWR
reactors and recent achievements in the field of nuclear power plant (NPP) safety. The main objective of the
project is to develop small, shipyard fabricated, multipurpose transportable NPP for safe operation over 10 to
12 years without refuelling at the berthing platform or on the coast. Plant maintenance and repair, refuellingand nuclear waste removal will be carried out at dedicated facilities.

2. Target Application
The ABV-6E RP is intended as a multi-purpose RP. The RP is designed with the capability of powering afloating power unit (FPU) as a part of floating nuclear power plant (FNPP) with a maximum length of 91.6 m,
a beam of 26 m, a draft of 3.6 m and a displacement of 8100 t. Depending on the needs of the region, the FNPP
can generate electric power or provide heat and power cogeneration or can be used for other applications.
Besides, a land-based configuration of the plant is also applicable.

3. Design Philosophy
The ABV-6E is a pressurized water reactor (PWR); its design incorporates the following main features:- Integral primary circuit layout with natural circulation of the primary coolant;- Negative feedbacks and enhanced thermal inertia;- Passive and self-actuated safety systems;- Increased resistance to extreme external events and personnel errors;- Use of nuclear fuel with the enrichment of less than 20%.
129

FPU includes reactor, steam-turbine, part of electric power system and control systems. The RPV operates
under conditions of 16.2 MPa in the reactor pressure vessel. The steam generators located inside the RPV
generate 295°C steam at 3.83 MPa flowing at 55 t/h. The RPV head is located under biological shielding andthe control rod drive mechanism is located above the shield outside the vessel.

4. Main Design Features
Reactor Core
The core comprises 121 hexagonal fuel assemblies (FA) of cassette type with active part height of 900 mm,
similar to the FAs in KLT-40S. Cermet fuel is used with less than 20% enriched 235U. Special stainless steel isused as fuel cladding.

Reactivity Control
Reactivity control without boron solution in the primary coolant and compensation of reactivity changes in
power operation is achieved by mechanical control and protection system (CPS). These inherent safety features
ensure automatic power regulation in a steady state operation, self-limiting power rise in case of positive
reactivity insertions, automatic control of the reactor power and primary coolant pressure and temperature intransients, as well as the emergency shutdown of the reactor core including the cases with a blackout and RPV
flip-over (with account of the time that the vessel flip-over process takes).

Reactor Pressure Vessel and Internals
The RPV is a welded cylindrical ‘container’ with an elliptical bottom. At the top of the vessel there are pipes
for feedwater supply and superheated steam removal, as well as those for the connection of the primary circuit
systems and the auxiliary process systems. The RPV head consists of a load-bearing slab, a shell attached tothis slab and sealed by a weld, and a top slab welded to the shell. The cavity between the top slab and the loadbearing slab is filled with serpentine which acts as a biological shielding, and the heat insulation is located at
the top. The posts of the CPS drives and thermal converters, etc. are welded to the load-bearing slab andpenetrate through the cover. Points of penetration through the top slab are sealed. Fuel assemblies are locatedin the in-vessel shaft. The protective tubes and devices provide the necessary coolant flow rate distribution
between the fuel assemblies and an arrangement of connectors for joining the absorber elements of fuelassemblies into CPS control rods and connecting the CPS control rods to CPS drives.

130

Reactor Coolant System
Core heat removal is based on conventional two-circuit methodology. The core is cooled and moderated bywater through natural circulation of coolant in the primary circuit. Hot coolant is cooled in a once-throughsteam generator, where slightly superheated steam is generated, then supplied to the turbine. This design
eliminates large-diameter pipelines in the primary circuit and main circulating pumps. The steam generator(SG), arranged in the annular space between the vessel and the in-vessel shaft, is a once-through verticalsurface-type heat exchanger generating steam of the required parameters from heat of the primary circuit
coolant. The SG is divided into four independent sections; feedwater supply and steam removal from eachsection is carried out through the pipes in the reactor vessel. Counter flow circulation is used, i.e., the primarycircuit coolant moves downward in the inter-tube space, while the secondary circuit coolant is moved upwardin the tubes. In case of inter-circuit leaks, it is possible to cut off any section automatically or remotely.
Identification of the leaking section is carried out with the use of the detection blocks of the radiation andprocess control system. Finding and disabling a faulty module is carried out during reactor shutdown.

5. Safety Features
Engineered Safety System Approach and Configuration
Safety of the ABV-6E RP is of utmost importance considering its close proximity to public area and at thesame time far-off location from main technical bases, which could provide timely technical support. In viewof its small power the emergency systems are simpler and often do not require active systems performance.
Land-based and floating power units use the advanced active and passive safety systems for emergency coolingover an unlimited time during design-basis and beyond design-basis accidents. Low thermal capacity of reactor
allows use of natural circulation in the primary coolant circuit and passive safety systems as primary safetysystems. The autoprotective features of the NPP have been improved for deployment in far flung territories.
The safety systems include:- Passive heat removal system;- Passive core cooling system;- Reactor caisson water flooding system;- Backup liquid absorber injection system

Decay Heat Removal System
In emergency modes, a combined-type residual heat removal system (RHRS) is used to remove decay heat.
This system functions on natural physical processes and - because there is an air heat exchanger cooled by theatmospheric air - ensures that the decay heat is being removed from the reactor for an unlimited time in all
types of accidents. Because of this, and considering the measures taken to enhance the reliability of the passive
RHRS, there are no active RHRS channels in the ABV 6E reactor design, which allows the output of emergency
power supply sources to be reduced. The passive RHRS is made of two independent channels connected to
two SGs each. Either channel, independently of the operability of the other channel, is capable of performing
the RHRS functions, i.e. of maintaining the parameters of the primary circuit in the design limits for an
unlimited time.

Emergency Core Cooling System
The emergency core cooling system (ECCS) is designed to compensate for the primary coolant leak and tocool the reactor core in case of LOCA. The ECCS comprises of the high-head pumps that inject water into the
RPV if power supply is available, and the hydro-accumulators that supply water under the action of thecompressed gas.

Containment System
The metal-and-water shielding (MWS) tank is a substantial structure for the equipment of the RP. RPV, twopressurizers and the cooler of the purification and heat removal system are enclosed inside the dry caissons of
the MWS tank. The passive reactor caisson water flooding system is designed to protect the RPV against meltdown in severe beyond-design-basis accidents associated with core damage. The system feeds the primarycoolant condensate to the RPV caisson. It is also possible to supply water from the fresh water intake andpumping system. The structure of the reactor caisson ensures the stable heat exchange between the RPV and
MWS tank.

6. Plant Safety and Operational Performances
The NPP with ABV-6E generates electricity and heat in the power range of 20–100%Nnom with the continuous
operation time of 26 000 hours. The NPP is designed for the manoeuvring rate of up to 0.1%/s. As a protection
against the external events, the NPP is equipped with both ground and waterside security structures. The
structures are designed for the sites in the Arctic zone with the frost penetration as deep as 2 m. The FPU and
NPP design is intended to withstand the 10-ton aircraft crash. As the analysis of emergencies has shown, theradiation and ecological impact to the personnel, public and the environment during normal operation,
abnormal operation, including the design-basis accidents, does not lead either to the excess of the radiation
doses established for the personnel and public, or release of any of radioactive content in the environment.
131This impact is also limited in beyond-design-basis accidents.

7. Plant Layout Arrangement

8. Design and Licensing Status
The final design of ABV-6E has been accomplished. The design has not been licensed yet.

9. Development Milestones
2006
2007
2014

132

Feasibility study developed for construction of the floating NPP with ABV-6M for the
Far North (settlement Tiksi, settlement Ust-Kamchatsk)
Feasibility study developed for construction of the floating NPP with ABV-6M for
Kazakhstan (City of Kurchatov)
Final design is being developed for a transportable reactor plant ABV-6E under thecontract with Minpromtorg (Russian Federation Ministry of Industry and Trade)RITM-200M (Afrikantov OKBM JSC,
Rosatom, Russian Federation)
All the content included in this section has been provided by and is reproduced with the permission of Afrikantov OKBM JSC, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

Afrikantov OKBM JSC, Rosatom,
Russian Federation
Integral PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary coolant system
circulation
NSSS operating pressure(primary/secondary), MPa
Core inlet/outlet coolant
temperature ( °C)
Fuel type/assembly array

175 / 50

Number of fuel assemblies in thecore
Fuel enrichment (%)

Forced circulation
15.7 / 3.83
282 / 318
UO2 (cermet fuel) pellets /hexagonal
241<20

Refueling cycle (months)

Up to 120

Core discharge burnup(GW·d/ton)Reactivity control mechanism–

Design service life (years)

Control rod drive mechanism;
within the control and protection
system (CPS)
Combined active and passive
system60

Ship footprint (m2)3360

RPV height/diameter (m)8.6 / 3.45

Approach to safety system

RPV weight (metric ton)265

Seismic design (SSE)

0.3g

Requirements for or approach tothe operating cycle
Distinguishing features

Without on-site refueling

Design status

Integral reactor, in-vessel corium
retention, double containment
6 prototype reactors were
manufactured and installed on
icebreakers (2 reactors are in theprocess of testing)

1. Introduction
RITM series reactors RITM-200 and RITM-200M are the latest development in Generation III+ SMR line
designed by the Afrikantov OKBM JSC and has incorporated all the best features from its predecessors.
Floating power units (FPUs) with the RITMs are commercially available for medium and long terms. The
RITM-200M adopts refueling cycle up to 10 years.

2. Target Application
The RITM-200M design was developed for the use on the Optimized Floating Power Unit (OFPU). The
OFPU is a power generating facility in the form of a compact non-self-propelled vessel having two RITM200M reactor plants. The FPUs based on RITM-200M may satisfy the needs of small residentials orindustrial facilities. The OFPU can provide electricity to domestic and industrial consumers. The OFPU canalso be used for heat supply and water desalination purposes after installing additional equipment.
133

3. Design Philosophy
RITM series reactors are the evolutionary development of the reactors (OK-150, OK-900, KLT-40 series) for
Russian nuclear icebreakers with a total operating experience of more than 60 years (more than 400 reactoryears). Incorporation the steam generators into the reactor pressure vessel, the reactor system andcontainment is very compact compared to the KLT-40. RITM design makes it possible to increase electric
output (40% more) and reduces the dimensions (45% less) and the mass (35% less) in compare with KLT40S.

4. Main Design Features
Nuclear Steam Supply System
RITM series nuclear steam supply system consists of the reactor core, four steam generators integrated in thereactor pressure vessel, four canned main circulation pumps (MCP), and pressurizer. The primary coolingsystem is based on forced circulation during normal operation and allows natural circulation for emergency
condition.

Reactor Core
RITM series adopts a low enriched cassette core similar to KLT-40S that ensures long time operation
without refueling and meets international non-proliferation requirements. The core consists of 241 fuelassemblies with the enrichment up to 20%. The core service life is up to 10 years.

Reactivity Control
Control rods are used for reactivity control. A group of control rods drive mechanisms is intended tocompensate for the excessive reactivity at start up, power operation and reactor trip. A group of shutdown
rods is designed for fast reactor shut down and to maintain it in the subcritical condition in case of accident.

Reactor Cooling System
The reactor pressure vessel (RPV) is thick-walled cylindrical pressure vessel. The reactor is designed as anintegral vessel with the main circulation pumps (MCP) located in separate external hydraulic chambers withside horizontal sockets for steam generator (SG) cassette nozzles. Each of the four SGs have 3 rectangularcassettes, while the four main circulation pumps are installed in the colds leg of the primary circulation pathand separated into four independent loops. The SGs generate steam of 295°C at 3.83 MPa flowing at 261(280) t/h. The conventional MCPs are used.

Steam Generator
The RITM uses once through (straight tube) SGs. The configuration of the steam generating cassettes makes
possible to compactly install them in the RPV.

Pressurizer
The design adopts pressure compensation gas system proved comprehensively in the Russian ship powerengineering. It is characterized by a simple design, which increases reliability, compactness, and no electric
power required. The design enables the use one pressurizer as a hydraulic accumulator, increasing reactor
plant reliability considerably in potential loss-of-coolant accidents.

5. Safety Features
The safety concept of the RITM is based on the defense-in-depth principle combined with the inherent safety
features and use of passive systems. RITM optimally combines passive and active safety systems to copewith abnormal operating occurrences and design basis accidents.
- Passive pressure reduction and cooling systems have been included (system reliability is confirmed by test
bench);- Pressure compensation system is divided into two independent groups to minimize size of potential coolant
leak;- Main circulation path of the primary circuit is located in a single vessel;- Steam header of primary coolant circulation is added, which ensures safety of the plant during SG and
MCP failures.

Approach to and Configuration of the Engineered Safety System
The high safety level of RITM series reactors is achieved both by inherent safety features and a combinationof passive and active safety systems. Moreover, redundancy of safety system equipment and channels andtheir functional and/or physical separation are provided to ensure high reliability. Safety systems are driven
automatically by the control system, when controlled parameters achieve appropriate set points. In case of
automated systems failure, self-actuating devices will actuate directly under the primary circuit pressure toensure reactor trip and initiate the safety systems.

134

Residual Heat Removal System
The residual heat removal system (RHRS) consists of four safety trains:- Active safety loop with forced circulation through steam generator.
- Active safety loop with forced circulation through primary-third circuit heat exchanger of primary circuit
coolant purification loop.
- Two passive safety loops with natural coolant circulation from water tanks through steam generators.
Evaporated is steam generators water condenses in air cooled heat exchangers and flow back to tanks withwater heat exchangers. After complete water evaporation from the tanks, the air cooled exchangers continue
provide cooling for unlimited time.

Emergency Core Cooling System
The ECCS consists of safety injection system (SIS) for water injection in primary circuit to mitigate theconsequences of a break loss-of-coolant accident. The system is based on active and passive principles withredundancy of active elements in each channel and consists of (i) two passive pressurized hydraulicaccumulators; (ii) two active channels with water tanks and two make-up pumps in each channel.

Containment System
RITM is placed within the hermetically sealed envelope with dimensions of 6.6 m×6.4 m×16.2 m around thereactor vessel to localize possible radioactive releases. In case of severe accident thick wall of the reactor
vessel keeps molten corium within the reactor. Water filled caisson under the reactor provides the reactor
vessel cooling. The envelope integrity ensured by overpressure relief valve, containment cooling system, andpassive autocatalytic recombiners.

6. Plant Safety and Operational Characteristics
The country of origin constructs the OFPU and conducts the first core loading. The transportation to theoperation site is through the territorial sea of transit countries. Power and heat are generated at the operationsite in the host country for up to 10 years without refueling. For maintenance, refueling and radwaste
handing, the OFPU is returned to the country of origin. Afterwards, it is transported to the operation site in
the customer’s country.

RITM-200M reactor safety system

OFPU life cycle

7. Instrumentation and Control System
An automated control system is provided in the RITM based nuclear power plant to monitor and control
plant processes. This system possesses necessary redundancy with regard to safety function fulfilment andallows both automated and remote control of the power plant.

8. General Layout of the Plant
The OFPU with the RITM-200M reactor is designed to supply electricity, thermal power, and desalinatedwater to coastal or isolated territories, offshore installations, islands, and archipelagos. The OFPU can be
rapidly delivered to the site by sea. The only needs to launch operation is docking pier and onshore powertransmitting infrastructure135

OFPU fitted with the RITM-200M reactor plant

9. Testing to Check and Validate the Design
The engineering solutions used in the design are traditional for marine power engineering. The solutions
have been checked in the course of many operating years and ensured the required reactor plant reliability
and safety performance. The RITM-200M reactor pertains to integral-type reactors. Integral-type reactors areused in a series of Project 22220 multipurpose nuclear-powered icebreakers Arktika, Sibir and Ural.

10. Design and Licensing Status
The RITM-series reactors have been developed in conformity with Russian laws, codes and standards in apeaceful use of atomic energy and in conformity with IAEA recommendations. At present, the reactors are
manufactured and installed on nuclear-powered icebreakers; the OFPU design is under development.

11. Approach to the Operating Cycle
The OFPU is delivered to the site with fresh fuel in its reactors. After the cycle of operation is over, the
OFPU comes back to the exporting country along with the spent fuel in its reactors. Spent nuclear fuel postirradiation handling and reprocessing are performed in the exporting country.

12. Waste Management System and Waste Disposal Plan
The waste is stored within the OFPU, not in the operation site water area. The waste ensuing from plant
operation is compact, has a low activity level and is securely isolated from the biosphere. It has been verified
that there is no effect to marine organisms in the deployment site water area.

13. Development Milestones
2016
2020
2020

136

The first RITM-200 installed on board the icebreaker Arktika
The icebreaker Arktika under testing
A conceptual design of the OFPU with RITM-200M

VBER-300 (JSC “Afrikantov OKBM”,
Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of Afrikantov OKBM, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

JSC “Afrikantov OKBM”,
Rosatom, Russian Federation
Integral PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

917 / 325

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

16.3

Number of fuel assemblies in thecore
Fuel enrichment (%)

85

Core Discharge Burnup (GWd/ton)50

Refuelling Cycle (months)

72

Reactivity control mechanism

Design life (years)

Control rod driving mechanismand soluble boron
Hybrid (active and passive)
system60

RPV height/diameter (m)9.3 / 3.9

Approach to safety systems

Forced circulation

292 / 328
UO2 pellet/hexagonal

4.95

Seismic Design (SSE)

0.25g

Distinguishing features

Power source for transportable
Floating NPPs, cogeneration
options, compact design
Licensing stage

Design status

1. Introduction
The VBER-300 is a multipurpose medium-sized power reactor with a rated electric power of 325 MW intended
for land-based nuclear power plants (NPPs), nuclear cogeneration plants, and transportable floating nuclear
power plants (FNPPs). The VBER-300 design is evolution of modular marine propulsion reactors. An increasein thermal power causes an increase in mass and overall dimensions; however, the reactor basic design issimilar to that of marine propulsion reactors. The VBER-300 design was developed based on the lessons
learned from the design, safety and operating experience for VVER reactors. VBER-300 adopts proven nuclear
ship building technologies and operating experience that in turn contribute to enhancement of operationalsafety and reduction in production costs. VBER-300 can be configured as a multi-module plant on request of
the customer. VBER-300 design features are availability for both land-based and transportable FNPPs, avariety of cogeneration options, maximally compact design, improved plant efficiency, and protection against
external impacts. A reduction in construction time is achieved due to the compact design of the reactor system.

2. Target Application
The VBER-300 nuclear plants are intended to supply thermal and electric power to remote areas where
centralized power is unavailable, and to substitute capacities of available cogeneration plants on fossil fuels.
The design is also proposed to be used as a power source for seawater desalination complexes. The VBER300 nuclear plant has two reactor units that operate in the steam-condensing mode and can generate 600 MW(e)to satisfy power demands of a city with a population of 300 000. According to the OKBM’s data, when VBER300 has cogeneration capabilities, the total electric output will reduce to 200 MW(e) providing 460 Gcal/hr forprocess heat applications.

137

3. Design Philosophy
VBER-300 design using ship-based modular configuration enhances the safety philosophy through provenmarine modular technologies. The reactor design has no pipelines in the primary circulation circuit. The reactor
unit incorporates the reactor and four steam generators – MCPs two-vessel units. The compact reactor systemcomprises the steam generating system in a limited space of the reactor compartment, and has enhanced
reliability and long refuelling cycle. VBER-300 can also be configured as a transportable FNPP and can bearranged to operate individually or as multi-module plant, increasing the power output by means of scaling up
the equipment and with the same reactor system configuration.

4. Main Design Features
Nuclear Steam Supply System
The separation heat exchangers are designed to extract heat energy from the nuclear heat source without mixing
the fluids circulating within the nuclear plant with those employed in the process heat application. In the
VBER-300 design, separation heat exchangers are thermally coupled indirectly via heat exchangers coupledwith the secondary loop supporting the power conversion system. In this configuration, a stream of the steam
generated via steam generators (for any of the 2, 3, and 4 SGs) and partially expanded in the turbines is
extracted at an intermediate pressure for circulation within the separation heat exchangers.

Reactor Core
The reactor core comprises 85 hexagonal fuel assemblies (FAs) which are placed in the reactor cavity in nodesof a regular triangular lattice with a space interval of 236 mm. Pelletized UO2 fuel with an enrichment of up
to 5% licensed and tested in VVER reactors is used. FA of unified design to increase the fuel efficiency isutilized. Each FA contains guide tubes that allow insertion/withdrawal of control rods. Reactor core also uses
gadolinium fuel elements, which contain gadolinium in the UO2 fuel pellet and has the same geometry as theregular fuel pellet.

Reactivity Control
Sixty one control rods in combination with fuel elements mixed with burnable poison materials provide safeand reliable reactivity control during both normal and transient operations. Control rods are operated throughhigh-performance electromechanical control rod drive mechanisms (CRDMs). The control rods elements aredesigned to maintain the core subcritical even if the most reactive assembly fails (i.e. stuck-rod/assembly
event). To compensate for the fuel burnup reactivity margin, fuel rods with gadolinium burnable poison
contained in uranium dioxide pellets are distributed across each FA with configurations similar to those usedin VVER-1000 reactors. Boric acid is also dissolved and maintained at controlled concentrations within the
primary coolant system to ensure optimum core power distribution.

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) consists of the reactor core and internals with an overall height of 9.3 m anda diameter of 3.9 m. The VBER-300 design provides a special system of emergency vessel cooling to solve
the problem of retaining the melt inside the reactor vessel in severe accidents. The core melt retention isfacilitated by the low power density, relatively low level of residual heat, no penetrations in the reactor vessel
bottom and smooth outer surfaces of the reactor vessel bottom creating more favourable conditions for steam
evacuation under core cooling by boiling water.

Reactor Coolant System
The VBER-300 primary cooling mechanism under normal operating conditions operates using forced
circulation of coolant by the MCPs and using natural circulation in the shutdown condition. The reliability andoperational safety of the MCPs are enhanced due to the usage of a proven technology and operating experience
for the pumps in the area of marine propulsion. The MCPs are connected directly to the steam generators(SGs). All components of the primary loop are directly connected to the RPV, except for the pressurizer. The
MCPs are centrifugal single-stage canned pumps with impellers.

Steam Generator
The SGs are once-through coil modules with the secondary coolant flowing inside the tubes. The feedwater ispumped through an inlet in the SG head, circulates within the SG tubes and exits through the SG outlet as asuperheated steam at the design pressure and temperature for expansion in the turbine generator units.

Pressurizer
The VBER-300 has an external steam pressurizer that is conventional for loop PWRs. The water region in thepressurizer, where electric heaters are located, is connected with the SG hot section in one primary loop. The
steam region of the pressurizer is connected with the cold section in this loop near the MCP pressure chamber,from which the underheated water is supplied to the pressurizer when valves are open in the injection line. The
pressurizer head in the steam region has two safety valves that protect the primary circuit against overpressure
in case of accidents with loss of decay heat removal.
138

5. Safety Features
The VBER-300 safety systems are based on the defence-in-depth principle with redundancy relying on
passively driven systems that enables the core to operate within safety margins under all anticipated accident
scenarios for at least 24 hours. After this initial period, emergency back-up and diverse safety systems ensure
continued core cooling for extended time. In addition, separation of the passive and active cooling channels
prevents common failures of the emergency core cooling systems (ECCS).

Engineered Safety System Approach and Configuration
The safety assurance and engineering solutions incorporated in the design focus on accident prevention
measures, design simplification, inherent safety; passive safety systems and enhancement of safety against
external events (including acts of terrorism); and mitigation of severe accident consequences. The RPV andconnecting piping that usually form the primary pressure boundary represent an additional physical barrier.
The leak-tight carbon steel containment and protective enclosure with filtration forms the ultimate barrier
separating the reactor system from the environment. For all cogeneration applications, the separation of heatexchangers represents a physical barrier to prevent radioactive release.

Decay Heat Removal System
The decay heat removal system (DHRS) consists of two passive heat exchangers and a process condenser.
Passive safety features are intended to arrange recirculation in the core for the removal of decay heat in thecourse of scheduled maintenance, refuelling or under emergency conditions. Passive emergency shutdown
cooling system operates using natural circulation of coolant in all heat transport circuits with stored watertanks, where water is evaporated and condensed back to liquid upon a contact with the cooler surfaces of thecontainment inner shell. Decay heat is also removed indirectly by the secondary circuit using the steam turbine
condenser.

Emergency Core Cooling System
The ECCS contains two stages accumulators with different flow-rate characteristics to ensure emergency corecooling for 24 hours, makeup pumps and a recirculation system. If electrical power is available during
accidents, makeup pumps and an active recirculation system ensures emergency core cooling beyond the initial
24 hours. The VBER-300 emergency shutdown system consists of the CRDMs, two trains of liquid absorber
injection, and two trains of boron control from the make-up system. Emergency residual heat removal system(RHRS) by means of passive cooling channels with water tanks and in-built heat exchangers, ensure reliablecooling up to 72 hours and longer. The system is actuated by passive means—hydraulically operated pneumatic
valves. The emergency core cooling accumulators are part of the passive water injection system as injection isdone using compressed gas. Containment depressurization systems prevent containment damage and reduceradioactive release in design basis accidents (DBA) and beyond DBAs. A small and medium loss of coolant
accidents (LOCA) are prevented by a combination of a sprinkler system, low-pressure emergency injection
system, and core passive flooding system.

Containment System
The land-based VBER-300 containment system includes a double protective pressure envelope formed by an
inner carbon steel shell and an outer reinforced concrete containment structure. In addition, localizing
reinforcement is provided to protect the pressure boundary represented by all auxiliary systems hydraulicallyconnected to the primary loop. The containment is designed to withstand all stressors induced by all credible
accident scenarios, including aircraft crashes. The inner steel containment of 30 m in diameter and 49 m highprovides space for condensing the steam generated from the medium in large LOCAs. The outer concrete
structure 44 m high and 36 m in diameter serves as protection against natural and man-caused impacts.

6. Plant Safety and Operational Performances
The VBER-300 safety concept is based on the defence-in-depth principles. With the modular configuration, it
has increased resistance to impact loads in case of earthquake and aircraft crash. Results of strength analysis
under seismic loads of up to 8 points according to the MSK-64 scale carried out for the VBER-300 RPs
confirmed that the reactor unit has a two-times safety margin (maximum seismic stress of the most loaded
vessel unit is 150 MPa maximum at an allowed stress of 370 MPa). Analysis of the 20 ton aircraft crash on thereactor compartment showed that overload upon the attachment fittings of the reactor unit is less than seismic
loads. It is considered that core melting accidents for the core are postulated. In the case of the severe accident,
the reactor cavity is filled with water from the emergency reactor vessel cooling system ensuring reliable heatremoval from the external surface of the bottom and lower portion of the vessel. Retention of satisfactorymechanical properties and load-carrying capacity of the vessel ensures retention of the melted core inside thereactor. The safety level of the power units with VBER plants correspond to requirements for Generation III+advanced nuclear stations making it possible to place them near cities that is of extreme importance as virtually
all regional power sources are used for district heating. The buffer area of the station coincides with theperimeter of the industrial site. The calculated radius of protection measures planning for population is 1 km.

139

7. Plant Layout Arrangement
In the basic architecture of the land-based VBER-300 power unit, the reactor, including its servicing systems,
spent fuel pool, and auxiliary equipment are arranged within a double containment resistant to aircraft crashes.

Reactor Building
The inner steel shell is a leak-tight cylindricalenclosure 30 m in diameter that is covered withthe semi-spherical dome 15 m in radius andthat has an elliptical bottom. The height of theleak-tight enclosure is 47 m. The steel shell isdesigned for parameters of the maximum DBA
with the excess pressure of 0.4 MPa and thetemperature of 150oC. The outer protective
enclosure is made of one-piece reinforced
concrete without preliminary tensioning of thesteel and consists of a cylindrical portion of thesemi-spherical dome. Building structures ofthe outer protective enclosure are designed forexternal accidental exposures, including anaircraft crash and air shock wave.

Balance of Plant
The VBER-300 design can be configured for land-based stationary applications wherein the reactor system –a nuclear island coupled to a turbine island and auxiliary buildings for spent fuel storage, water treatment,
maintenance, and switchyard connections with configurations similar to conventional large LWRs – are housed
in a relatively small area.

Turbine Generator Building
Each VBER-300 reactor system can be thermally coupled with one or multiple turbine generator sets. A
slightly superheated steam is supplied to the turbine in the secondary circuit with part of the steam taken off
from the turbine and directed to the heat exchanger of a district heating circuit. It can operate as a NPP with acondensing turbine and as a nuclear cogeneration plant with a cogeneration turbine.

8. Design and Licensing Status
Development of the final design and design documentation for a VBER-300 nuclear station can begin
immediately upon the request of a customer. It will take 36 months to develop documentation to the extent
needed to obtain a license for VBER-300 NPP construction, including 18 months to develop the design.

9. Fuel Cycle Approach
The VBER-300 design concept allows a flexible fuel cycle for the reactor core with standard VVER FAs. The
fuel cycles are 3×2 years and 4×1.5 years. The number of FAs in the refuelling batch is either 15 or 30; maximal
fuel burnup does not exceed 60 GWd/ton U for the cycle with 30 fresh FAs in the reloading batch andmaximum initial uranium enrichment.

10. Development Milestones
2001
2002
2004
2006
2007–20092007–2008
2008–20112011–20122012–2015

140

Design activities to develop VBER reactors started
Technical and commercial proposal for the two-unit VBER NPP
Preliminary design 1 approved by the Scientific and Technical Board and State Nuclear
Supervision Body (GosAtomNadzor)
JSC “Kazakhstan-Russian company “Atomic stations” was established to promote the
VBER-300 design.
Technical Assignment for the NPP design and final designs of the reactor plant,
automated process control system, and heat-generating plant; feasibility, economy, andinvestment studies of the VBER-300 RP NPP for the Mangistau Region, Kazakhstan
Development of the 100–600 MW VBER plant
R&D for the VBER-460/600 NPP design
Development of the VBER-600/4 NPP based on the heat exchange loop of the increased
capacity
Technical and economic optimization of the VBER-600/4 plant

SHELF-M (NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of, NIKIET, Russian Federation.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
NIKIET, Russian Federation
country of origin
Reactor type

Integral PWR

Coolant/moderator
Thermal/electrical
capacity, MW(t)/MW(e)Light water / light water
35.2 / up to 10

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Natural (for 20 % of rated power) / forced
14.7 / 3.7

Core inlet/outlet coolant
temperature (oC)

271 / 308

Fuel type/assembly array

Cermet (UO2 + metallic silumin) /hexagonal

Number of fuel assemblies
in the core

241

Fuel enrichment (%)
Refuelling cycle (months)

Up to 19.7
96Core discharge burnup(GWd/ton)

162.4

Reactivity control
mechanismApproach to safetysystems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)Control rods driving mechanism

RPV weight (metric ton)

Power capsule: 350
RPV: 75

Seismic design (SSE)
Fuel cycle
requirements/approach

VIII by MSK-64 scale
Once-through fuel cycle

Distinguishing features

Autonomous passive decay heat removal
system, integral transportablecontainment

Design status

Basic design

Combined active and passive
60
16 800Power capsule: 10 / 8
RPV: 4 / 2.6

1. Introduction
The SHELF-M is a modernized NPP based on SHELF reactor unit with an increased power capacity of 35.2
MW(t) and a rated power output of up to 10 MW(e) developed based on NIKIET's experience in marine nuclear
reactors design. The plant modular design and small containment vessel makes it possible to deliver reactor
unit on site fully assembled and tested, minimizing construction costs and time. Refuelling and nuclear waste
removal will be carried out at dedicated facilities. Optimized core design allowed to increase refuelling cycle
for up to 8 years.

2. Target Application
The SHELF-M is a land-based NPP intended to supply electric power to remote areas with decentralized power
supply. The plant does not require local water sources - residual and decay heat is removed by external heatexchangers cooled by atmospheric air. The SHELF-M is designed to be operated with low staffing level with
only 15 staff members required to be present of site during normal operation.

3. Design Philosophy
The SHELF-M is based on proven technology derived from NIKIET's experience in design and development141

of marine propulsion PWR with the addition of advanced passive safety systems. Integral RPV layout with a
combined forced and natural primary coolant circulation makes it possible to achieve significant reduction in
power unit weight and size. RPV and all necessary normal operation and safety systems are assembled inside
a cylindrical power capsule with 8 m in inner diameter and 10 m long. Capsule's shell serves as a containment.
Compared to SHELF, SHELF-M has optimized core with extended fuel cycle (up to 8 years) and higher
thermal capacity (35.2 MW).

4. Main Design Features(a) Nuclear Steam Supply System
The SHELF-M power unit implements a traditional two-circuit NSSS. SG provides overheated steam to two
turbines located in the engine room adjacent to power capsule room and connected to power capsule with 4
independent steam pipelines. Each pipeline is equipped with shut-off valves installed both inside and outsideof power capsule and can be independently isolated without reactor shutdown in case of leaks of primarycoolant into the secondary coolant.
(b) Reactor CoreThe SHELF-M reactor core consists of 241 cylindrical fuel assemblies (FA) of different types with variable
fuel enrichment. Fuel assemblies are placed in nodes of a regular hexagonal lattice. Control rods are locatedbetween FA and united in 8 groups. Each fuel assembly contains 55 fuel rods with cermet fuel placed in nodesof regular hexagonal lattice and 6 stationary rods of burnable absorber around fuel rods. Emergency protection
rods are located inside 18 fuel assemblies in place of 19 central fuel rods. Cermet fuel consists of uranium
dioxide fuel particles embedded in silicon-aluminium (silumin) matrix. Fuel rod shell is made of chromium
nickel alloy. High thermal conductivity of fuel matrix as well as absence of zirconium in structural elementsof fuel rod improves design safety in transients and prevents hydrogen production in steam oxidation of
zirconium alloys.
(c) Reactivity Control
The reactor core contains two independent reactor shutdown systems. Control rods with boron carbide absorberat the top and dysprosium titanate at the bottom provide normal operation reactivity control and shutdown.
Emergency protection rods with boron carbide ensure reliable shutdown in case of emergency. The SHELFM is also equipped with ultimate shutdown system — emergency liquid absorber injection system (ELAIS),
which is able to shutdown the core in case of severe accident by injecting boron acid in primary coolant.
(d) Reactor Pressure Vessel and Internals
The RPV accommodates the core with reactivity control rods, the primary coolant circuit, the mechanical filter,
the thermal shielding, the steam generator and heat exchangers of emergency decay heat removal system. The
RPV has an elliptical bottom, cylindrical shells and two rector covers (central and peripheral). All pipelines
are connected to the covers at the top of the RPV to prevent the core drainage in case of LOCA.
(e) Reactor Coolant System
Reactor coolant system is based on conventional two-circuit methodology. The SHELF-M primary circuit
cooling under normal operating conditions is done using forced circulation by two primary coolant circulation
pumps (PCCP). Cooling by natural circulation is also possible with thermal output at the level 20 % of rated
thermal power and during shutdown.
(f) Steam Generator
The SHELF-M utilises once-through steam generator with helically coiled tubes located inside the cylindricalannulus between the RPV and the core barrel. The SG comprises a tubing, collection and distribution chambers
above and below the tubes, steam and feedwater lines inside the RPV. SG coil-pipes, alongside with emergency
decay heat removal system (EDHRS) heat exchanger (HX) coil-pipes form a single coil with 48 independent
modules grouped into 8 sections: 4 for SG and 4 for EDHRS HX, which makes it possible to cut off individual
section in case of primary coolant leaks.
(g) Pressurizer
The SHELF-M adopts pressure compensation gas system common in Russian marine PWR. It utilizes anexternal pressurizer with no active components, such as sprinkler system or electrical heating. Pressurizer
system consists of 5 separate vessels: gas collector, intermediate pressurizer, end pressurizer and two
expansion vessels. Pressurizer is connected to the reactor central cover by a pipeline with a restrictor to reduce
coolant leak rate in large-break accidents.
(h) Primary pumps
The SHELF-M utilizes two electrical primary coolant circulation pumps (PCCP) located on peripheral coverof the RPV.

142

5. Safety Features(a) Engineered Safety System Approach and Configuration
The SHELF-M is designed with combined passive and active safety systems that comprises emergency decay
heat removal system (EDHRS), emergency reactor cooling system (ERCS), emergency liquid absorber
injection system (ELAIS) and overpressure protection systems.
(b) Decay Heat Removal System
During normal shutdown, decay heat is removed from the reactor core by forced circulation of primary coolant
to SG and then by secondary coolant to the condenser. In case of emergency (primary or secondary circuit
pumps failure, secondary circuit loss of coolant etc), emergency decay heat removal system (EDHRS) is used
to cool down the core. The system consists of 4 independent loops with two passive circuits in each loop
(intermediate circuit and low-boiling circuit) with natural circulation of coolants and stand-alone heatexchangers between circuits. The heat is transferred to the atmospheric air via heat exchangers, located on theroof of the reactor building.
(c) Emergency Core Cooling System
The emergency reactor cooling system is designed to supply the in-vessel circulation circuit with water during
accidents with loss of the primary circuit integrity. The system is based on passive principals and doesn’t
require any manual or automatic activation. The system contains 4 vessels with coolant under high pressure,which is passively injected into RPV in case of primary coolant pressure drop.
(d) Containment System
The SHELF-M utilizes two containment systems. The first one — safety vessel, located inside power capsule,that encloses RPV and all primary circuit equipment. The second — power capsule shell, which contains safety
vessel with RPV, steam and feedwater pipelines, EDHRS intermediate heat exchangers, reactor unit auxiliary
equipment and tanks with liquid and solid radioactive waste. Both containment system are designed towithstand pressure and temperature of primary coolant in case of large-break accidents. Both containments areunvisited during normal operation and filled with inert gas (nitrogen) with absolute pressure below atmospheric
to prevent minor leaks of radioactive media and reduce the probability and possible consequences of fire insidecontainment volume. Both containments are equipped with overpressure protection systems.

6. Plant Safety and Operational Performances
The electric power of a single SHELF-M unit is up to 10 MW(e), and the thermal power is 35.2 MW(t). The
current supplied to the consumer system is alternate and three-phase with voltage 0.4 kV ± 2 %, frequency 50
Hz ± 1 Hz). The NPP base operation mode is power operation in range from 20 to 100 % of rated power withthe capability to vary the consumed power daily and annually. The power increase and decrease rate is 1 %(with forced primary coolant circulation). Design capacity factor - 0.9. Refuelling outage time is 30 days oncein 8 years.

7. Instrumentation and Control System
The automated process control system (APCS) of a NPP with SHELF-M reactor unit is to control major andauxilary electricity generation processes in all modes of the unit operation:
1. Normal operation comprises of phased automated initiation, operation at steady power levels in a range of
20 to 100 % of rated power with forced primary coolant circulation, operation at steady power levels at 20 %of rated power with natural coolant circulation, switchover from natural primary coolant circulation to forced
circulation and scheduled automated deactivation.
2. Anticipated operational occurrences like emergency power reduction and operation with a decreased steam
supply due to failures of the reactor facility's key components or feedwater supply and steam receipt systems.
3. Emergency: emergency deactivation in the event of reactor facility parameters deviation beyond the safe
operation limits or in the event of equipment failures leading to the safe operation limits being violated.

8. Plant Layout Arrangement
The SHELF-M plant consists of main building, administrative building, 4 stand-alone fan cooling towers, 4
emergency diesel generators, start-up boiler module, water storage tanks, fresh and spent fuel open storage
area and auxiliary modules. The main building is based on lightweight steel frame structure and houses thereactor hall with power capsule and external concrete protection, and the turbine hall. The building is
48×48×25 m is size and is designed as quick assembly building based on prefabricated structures, that allows
to significantly reduce construction costs and time.
The administrative building is assembled from standard containers and houses administrative offices andcontrol room.

9. Testing Conducted for Design Verification and Validation
Experimental activities are built and in operation. The SHELF-M reactor unit fuel rods reactor tests are under
way.
143

10. Design and Licensing Status
Licensing process to be started in 2025.

11. Fuel Cycle Approach
The SHELF-M implements a standard PWR fuel cycle. All fuel assemblies are discharged at the end ofrefuelling cycle. On-site (for FOAK plant) and remote (for subsequent plants) refuelling strategies are designed.
For the FOAK plant, the fresh and spent open air fuel storage is located on site.

12. Waste Management and Disposal Plan
Liquid radioactive waste management system is located inside power capsule. The system is designed to gather
radioactive leaks from reactor cooling systems and storage liquid waste inside power capsule.
Gaseous radioactive waste system is designed to remove gaseous radioactive waste from within the power
capsule during the scheduled maintenance (one a year). Nitrogen from within power capsule passes throughfilters, where radioactive aerosols are absorbed and stored.

13. Development Milestones
2018
2019
2023
2025
2030

144

Start of the project SHELF-M (achieved)
Conceptual design (achieved)
Preliminary design (ongoing)
Licensing, detailed design, start of construction
Operation testingPART II. HIGH TEMPERATUREGAS COOLEDSMALL MODULAR REACTORS

145

146

HTR-PM (Tsinghua University, China)
All content included in this section has been provided by, and is reproduced with permission of Tsinghua University, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Technology developer, country oforiginReactor type

Value
INET, Tsinghua University,
China
Modular pebble bed HTGR

Coolant/moderator

Helium/graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

2 × 250 / 210

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array7 / 13.25

Number of fuel assemblies in thecore
Fuel enrichment (%)

Forced circulation

250 / 750
Spherical elements with coatedparticle fuel420 000 (in each reactor module)
8.5

Core Discharge Burnup (GWd/ton)90

Refuelling Cycle (months)

On-line refuelling

Reactivity control mechanism

Control rod insertion

Approach to safety systems

Combined active and passive

Design life (years)

40

Plant footprint (m2)

256 100

RPV height/diameter (m)25 / 5.7 (inner)

RPV weight (metric ton)

800

Seismic Design (SSE)

0.2 g

Fuel cycle requirements / Approach

LEU, open cycle, spent fuelintermediate storage at the plant
Inherent safety, no need foroffsite emergency measures
In operation

Distinguishing features
Design status

1. Introduction
In 1992, the China Central Government approved the construction of the 10 MW(t) pebble bed hightemperature gas cooled test reactor (HTR-10) in Tsinghua University’s Institute of Nuclear and New Energy
Technology (INET). In 2003, the HTR-10 reached its full power operation. After that, INET has completedmany experiments on the HTR-10 to verify crucial inherent safety features of modular HTRs, including (i)loss of off-site power without scram; (ii) main helium blower shutdown without scram; (iii) withdrawal ofcontrol rod without scram; and (iv) helium blower trip without closing outlet cut-off valve.
The next step of HTR development in China began in 2001 when the high-temperature gas-cooled reactor-pebble-bed module (HTR-PM) project was launched. The first concrete of the HTR-PM demonstration power
plant was poured on 9 December 2012, in Rongcheng, Shandong Province. In support of manufacturing firstof a kind equipment and licensing, large scale engineering test facilities were constructed, and all tests havebeen completed. The civil work of the nuclear island’s buildings has been completed in 2016 with the first of
two reactor pressure vessels installed in March 2016. Currently all major equipment has been manufactured
and already installed. The power plant is scheduled to start power generation in 2021.

2. Target Application
The HTR-PM is a commercial demonstration unit for electricity production. The twin reactor modules drivinga single turbine configuration was specifically selected to demonstrate its feasibility. Following the HTR-PM
demonstration plant, commercial deployment of HTR-PM based on batch construction is planned. Units withmultiple standardized reactor modules coupling to one single steam turbine, such as 200, 600 or 1000MW(e)147are envisaged.
A standard design has been finished for the 600 MW(e) multi-module HTR-PM600 nuclear power plant, which
consists of six reactor modules or Nuclear Steam Supply System (NSSS) modules coupling to one steam
turbine. Each NSSS module has the same design as the HTR-PM demonstration plant, with independent safetysystems and shared non-safety auxiliary systems. The footprint of a multi-module HTR-PM600 plant is notsubstantially different from that of a PWR plant generating the same power. Future sites have been identified
for possible deployment. There is a big market demand for high temperature steam supply from HTRPM/HTR-PM600 in China.

3. Design Philosophy
The HTR-PM consists of two NSSS modules coupled with a 210 MW(e) steam turbine, as shown below. Each
NSSS module includes a reactor that contains reactor pressure vessel, graphite, carbon, and metallic reactor
internals; a steam generator; a main helium blower; and a hot gas duct. The thermal power of each reactor
module is 250 MW(t), the helium temperatures at the reactor core inlet/ outlet are 250/750°C, and steam
parameters is 13.25 MPa/567°C at the steam turbine entrance.

4. Main Design Features
Reactor Core and Power Conversion Unit
The primary helium coolant works at 7.0 MPa with the rated mass flow rate of 96 kg/s. Helium coolant enters
the reactor in the bottom area inside the RPV with an inlet temperature of 250ºC. Helium coolant flows upwardin the side reflector channels to the top reflector level where it reverses the flow direction and flow into thepebble bed in a downward flow pattern. Bypass flows are introduced into the fuel discharge tubes to cool thefuel elements there and into the control rod channels for control rods cooling. Helium is heated up in the active
reactor core and then is mixed to the average outlet temperature of 750ºC and then flows to the steam generator.

Fuel Characteristics
Illustrated above, fuel elements are spherical ones. Every fuel element contains 7 grams of heavy metal. The
enrichment for the equilibrium core is 8.5% of 235U. Uranium kernels of about 0.5 mm diameter are coated bythree layers of pyro-carbon and one layer of silicon carbon. Coated fuel particles are dispersed in matrix
graphite, 5 cm in diameter. Surrounding the fuel containing graphite matrix is a 5 mm thick graphite layer.

Fuel Handling System
The operation mode of HTR-PM adopts continuous fuel loading and discharging: the fuel elements drop into
the reactor core from the central fuel loading tube and are discharged through a fuel extraction pipe at the corebottom. Subsequently, the discharged fuel elements pass the burn-up measurement device one by one. Whena fuel sphere reaches the target burnup they will be discharged into the spent fuel storage tank, otherwise they
are re-inserted into the reactor to pass the core once again.

Reactivity Control
Two independent shutdown systems are installed: a control rod system and a small absorber sphere (SAS)system, both placed in holes of the graphite side reflector. Reactivity control is performed using 24 control rod
assemblies, and 6 SAS shutdown systems serve as a reserve shutdown system. The control rods are used as aregulating group during normal plant operation and for emergency shutdown. Furthermore, turning off thehelium circulator is also efficient for reactor trip. Drop of all control rods can achieve long term shutdown.
The SAS system is used to reduce the shutdown temperature for the purpose of in-service inspection andmaintenance. Absorber material of control rods and small absorbers is B4C.

Reactor Pressure Vessel and Internals
The primary pressure boundary consists of the reactor pressure vessel (RPV), the steam generator pressure
vessel (SGPV) and the hot gas duct pressure vessel (HDPV), which all are housed in a concrete shielding
cavity. The three primary pressure vessels are composed of SA533-B steel as the plate material and (or) the
508-3 steel as the forging material. The ceramic structures surrounding the reactor core consist of the innergraphite reflector and outer carbon brick layers. The whole ceramic internals are installed inside a metallic
core barrel, which itself is supported by the RPV. The metallic core barrel and the pressure vessel are protectedagainst high temperatures from the core by the cold helium borings of the side reflector, which act like ashielding temperature screen.
148

5. Safety Features
The HTR-PM is designed with the following safety features: (1) radioactive inventory in the primary helium
coolant is very small during normal operation conditions, and even if released there is no need to take any
emergency measures; (2) for any reactivity accident or loss of coolant accident, the rise of the fuel elements’temperature will not cause a significant additional release of radioactive substances; (3) the consequences ofwater or air ingress accidents depend on the quantity of such ingresses. The ingress processes and theassociated chemical reactions are slow, and can readily be terminated within several dozens of hours (or evensome days) by taking very simple actions. And air ingress accident is classified as design extension condition.
The HTR-PM incorporates the inherent safety principles of the modular HTGR. The lower power density,
good coated particle fuel performance and a balanced system design ensures that the fundamental safetyfunctions are maintained. A large negative temperature coefficient, large temperature margin, low excess
reactivity (due to on-line refuelling) and control rods ensure safe operation and limit accident temperatures.
The decay heat is passively removed from the core under any designed accident conditions by natural
mechanisms, such as heat conduction or heat radiation, and keeps the maximum fuel temperature below
1620°C, so as to contain nearly all of the fission products inside the SiC layer of the TRISO coated fuel particles.
This eliminates the possibility of core melt and large releases of radioactivity into the environment.
Another feature of the HTR-PM design is the long-time period of accident progression due to the large heat
capacity of fuel elements and graphite internal structures. It requires days for the fuel elements to reach the
maximum temperature when the coolant is completely lost.

Engineered Safety System Approach and Configuration
When accidents occur, a limited number of reactor protection actions shall be called upon by the reactor
protection system. No or very limited actions through any systems or human interventions are foreseen after
the limited reactor protection actions are activated. The limited reactor protection actions shall be to trip thereactor and the helium circulator, to isolate the primary and secondary systems. When there is large leak or
rupture of steam generator heat transfer tubes, a dumping system is designed to minimize the amount of water
ingress into the primary circuit.

Reactivity control
The on-line refuelling leads to a small excess reactivity, the overall temperature coefficient of reactivity is
negative, and two independent shutdown systems are available.

Reactor Cooling Philosophy
Normally the reactor is cooled by steam generating system. Under accident conditions, the main helium blower
shall be stopped automatically. Because of the low power density and the large heat capacity of the graphite
structures, the decay heat in the fuel elements can dissipate to the outside of the reactor pressure vessel bymeans of heat conduction and radiation within the core internal structures, without leading to unacceptable
fuel temperature. And the fuel temperature increase in this phase will compensate accident reactivity andshutdown the reactor automatically via negative temperature feedback. The decay heat shall be removed toheat sink passively by reactor cavity cooling system (RCCS). Even if the RCCS fails, the decay heat can beremoved by transferring it through the concrete structure of reactor cavity while the temperatures of fuelelements are under design limit.

Containment Function
Retention of radioactivity materials is achieved through multi-barriers. The fuel elements with coated particles
serve as the first barrier. The fuel elements used for HTR-PM have been demonstrated to be capable ofretaining fission products within the coated particles under temperatures of 1620ºC which is not expected forany plausible accident scenarios. The second barrier is the primary pressure boundary which consists of thepressure vessels of the primary components. The vented low-pressure containment (VLPC) is designed
according to ALARA principle to mitigate the influence of accidents, consisting of the reactor cavity inside
reactor building and some auxiliary systems such as sub-atmosphere ventilation, burst disc and filters.

Chemical control
Water and steam ingress is limited by plant design (pipe diameters, SG lower than core and water / steam
dumping system) while massive air ingress is practically eliminated (small pipes, connection vessel; nochimney effect). After water ingress accident, the way to remove humidity from primary circuit are provided.

6. Plant Safety and Operational Performances
The HTR-PM demonstration power plant is under final commissioning test phase. Due to online fuel loading
mode of HTR-PM, better availability factor can be expected compared with other power plants operating in amode of periodic fuel loading.

7. Instrumentation and Control Systems
The instrumentation and control system of HTR-PM is similar to those of normal PWR plant. The two reactor
modules are controlled in a coordinated manner to meet various operational requirements.
149

8. Plant Layout Arrangement
The nuclear island contains reactor building, nuclear auxiliary building, spent fuel storage building and Electricbuilding, as shown below. The steam turbo-generator, which is similar to that of a conventional fossil-fired
power plant, is housed in the turbine building.

9. Design and Licensing Status
The preliminary safety analysis report (PSAR) was reviewed by the licensing authorities during 2008-2009.
The Construction Permit was issued in December 2012. Final approval of the FSAR is achieved in July 2021,
and Operation License was obtained in August 2021, followed by fuel loading, criticality and power operation.

HTR-PM Main Control Room

10. Fuel Cycle Approach
The air-cooled spent fuel canisters are placed in the spent fuel storage building with concrete shields. The
canister can be placed in a standard LWR transport cask and be transported if necessary. HTR-PM currently
adopts open fuel cycle. After the intermediate storage of spent fuel elements, the final storage in geologicaldeposits can be carried out in the open cycle. In the closed cycle, the spent fuel elements would be dismantled,
and the nuclear fuel can be reprocessed in normal reprocessing facilities (when the amount of spent fuel reaches
certain level and reprocessing technology is economically available).

11. Waste Management and Disposal Plan
The technologies of cleaning the liquid waste and of the off-gases are similar to those used in normal PWR
plants, although the amount of liquid waste from HTR-PM is much smaller. Waste with low or medium level
activity resulting from the operation is conditioned following different process technologies, which have been
established with high efficiency in nuclear industry. Different auxiliary material and solid residue will be put
into casks for intermediate storage. These waste can be underwent a final storage after conditioned.

12. Development Milestones
2001
2004
2006
2006
2006-2008
2009
2012
2013
2014
2015
2016
2017
Q4/20202021.82021.92021.12.20
2022

150

Launch of commercial HTR-PM project
Standard design of HTR-PM started
HTR-PM demonstration power plant approved as one of National Science andTechnology Major Projects
Huaneng Shandong Shidaowan Nuclear Power Co., Ltd, the owner of the HTR-PM,
established by the China Huaneng Group, the China Nuclear Engineering Group Co. andTsinghua University
Basic design of HTR-PM completed
Assessment of HTR-PM PSAR completedFirst Pour of Concrete of HTR-PM
Fuel plant construction started
Qualification irradiation tests of fuel elements completed
Civil work of reactor building finished
RPV and core barrel etc. delivered, installation of main components ongoing
Fuel plant achieved expected production capacity
Startup commissioning test of primary circuit
Operation License
First criticality
Grid connection
Full power operation

STARCORE (StarCore Nuclear, Canada,
United Kingdom and United States)
All content included in this section has been provided by, and is reproduced with permission of StarCore Nuclear, Canada

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

StarCore Nuclear, Canada

Coolant/moderator

Helium/Graphite

Thermal/electrical capacity,
MW(t)/MW(e)Block One: 35 / 14
Block Two: 50 / 20
Block Three: 150 / 60
One to six modules/plant
ForcedPrimary circulation

HTGR

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

7.4 / 6.7

Number of fuel assemblies in the core
Fuel enrichment (%)

Block 1: 90; Block 2: 126; andBlock 3: 324
15Core Discharge Burnup (GWd/ton)

60 GWd/t / 6 %Refuelling Cycle (months)> 60

Reactivity control mechanism

Design life (years)

2 x 6 Rotary/Automatic
deployment when gas flow lost
Passive; Containment 1: TRISO;
Containment 2: RPV/ETS-1
Helium System;
Containment 3: Silo
40 – 60

Plant footprint (km2)

1 (up to 1 Hectare)

RPV height/diameter (m)Block One and Two: 7.4 / 2.6
Block Three: 9 / 3.2
Block One: 75; Block Two: 90;
Block Three: 150
0.25g

Approach to safety systems

RPV weight (metric ton)
Seismic Design (SSE)
Fuel cycle requirements / ApproachDistinguishing features
Design status

280 / 750
TRISO Prismatic

LEU/Temporary storage in Silo atplant
RPV 30 Metres below grade in
hardened silos
Pre-conceptual / conceptual
design1. Introduction
Founded in 2008, StarCore had the first major Technical Review at Argonne National Laboratory in 2013, andthe last major Due Diligence (technical and business) review in 2019. StarCore is teamed with TREDIC
Corporation (UK), a global infrastructure developer to export plants from Canada to all compliant Nations
around the Globe. StarCore provides risk free, inherently safe, Generation IV Small Modular Reactor (SMR)
power to off-grid and edge-of-grid locations. Our operations can provide clean electricity for industrial andconsumer use, high temperature thermal energy for down-stream minerals processing, desalinated or purified
water for irrigation or for those people without clean water sources, and wide-band internet for medical andeducational use.
StarCore’s business model is predicated on a ‘Build, Own, Operate and Decommission’ (BOOD) basis,
generating revenue and profit from Power Purchase Agreements (PPAs) and off-take arrangements with bothnational and local governments, and private enterprise. The StarCore business plan therefore brings with it a151

very low financial burden and StarCore will be responsible for obtaining all the necessary licenses andcertification.

2. Target Applications
Serving remote communities to provide energy access and combat energy poverty is an important part of
StarCore’s vision. In the developing countries there are more than two billion people who are either without
electricity or who get power from diesel generators at very high cost. Many countries lack supplies of cleanwater and desalination-only plants are attractive to them. We are aware that we cannot just bring energy to aremote community – we must also bring some type of environmentally suitable industry, so the people have afuture. We include 100 km HVDC transmission lines in our cost estimates, and so we can connect plants toresidential and manufacturing sites. Remote mines are also an ideal market for StarCore, since they tend to be
in remote locations and energy costs are high.

3. Design Philosophy
StarCore is designed to operate in the harshest environment in remote locations anywhere in the world. To thisend the reactors are contained in steel-walled concrete silo structures 30 metres below grade. Each reactor silo
has two supporting silos with their access hatches in the refuelling room, also below grade. This room is filledwith helium at low pressure during refuelling operations, which are carried out by an automatic system.
StarCore plants have between one and six reactors per plant, giving us a range of outputs between 14 MW(e)and 360 MW(e). The plants are load-following, and can also output High Temperature Gas or Steam, with theoutputs dynamically changed as required.
The plants meet all planned Remote Siting Requirements, which include that they must be: Inherently Safe;
Passively Secure; Load Following; Fully Automated; Have a Remote Shutdown (Intervention) Capability; andafter suitable qualification (subject to local regulations) be operated with a zero-radius exclusion zone.
The plants also have 69 KV 100 km HVDC transmission lines included in the cost of the plant. The StarCore
Build-Own-Operate-Decommission Business Plan includes all capital costs, licensing, operating anddecommissioning logistics, thus eliminating the cost of entry to all countries and communities. The HVDC isinverted to local frequency standards.

4. Main Design Features
Reactor Core
The prismatic core is made up of hexagonal graphite blocks that are 360 mm across the flats and 793 mm long.
Cylindrical fuel compacts (26 mm diameter and 39 mm long) are inserted into holes drilled in the graphite
blocks, and burnable (neutron) poison elements will also be inserted as needed. The helium flow will bethrough vertical holes drilled in the blocks. The number of blocks used depend on the output of the core.
The helium coolant enters at the bottom of the core, flows up the outside including through the reactivity
control mechanisms, and then down through the core prismatic blocks. There are automatic bi-stable valves at
the inlet and outlet, which seal off the core in the event of pressure loss.

Reactivity Control
The core has vertical rotary reactivity control rods with a neutron reflector on a semi-circular rotating halfcylinder. There are two sets of 6 control rods each with a different deployment mechanism that deploys thecontrols if helium flow or pressure is lost. The reactivity control mechanisms use the helium pressure
differential across the core to stay stowed; in the event of pressure or flow loss they automatically deploy.

Reactor Pressure Vessel and Internals
The primary containment boundary is the TRISO fuel microsphere; the second is the Energy Transport System(ETS-1) with the nuclear system containing the RPV, piping, pumps and first stage intermediate heatexchanger. IHX-1. There are helium scrubbers in ETS to remove trace elements of radioactive dust and tritium.

Reactor Coolant System
The first stage coolant Energy Transport System (ETS-1) is helium at 7.3 MPa; this then transfers energy to
ETS-2 through IHX-1 which contains nitrogen at 6.7 MPa. Nitrogen is used in this system at this pressure tominimize delta-P across IHX-1 and allow compact exchanger designs to be sued; in addition, the pressure
gradient ensures that any gas migration will be in the direction of ETS-2.
The Nitrogen in ETS-2 then passes to an aero-derivative gas turbine that has an annular heat exchanger inplace of the usual burner cans, and the turbine exhaust gas at 300 oC can be used for district heating or cooling.
The energy in ETS-2 can also be sent to a heat exchanger that provides high-temperature gas or steam toexternal plants.

5. Safety Features
The TRISO fuel exhibits a very strong negative temperature coefficient. As the fuel temperature increases theneutron energy also increases; this effect reduces the neutron cross sections and lowers the number of fissions
and thus the power level. The result of removing all reactivity controls and shutting off the primary andsecondary cooling systems will result in the core becoming stable with an output of 600 kW(t), which will be152dissipated to the outside of the silo through vestigial fins into iso-thermal layer of the surroundings. The
prismatic core is attractive in this regard since the thermal pathways are better than those in pebble bed cores.
StarCore also has automatically deployed reactivity controls and inlet and outlet shutoff valves that are
deployed if helium pressure or flow is lost. In a worst-case accident, where all control mechanisms fail, andthe helium is exhausted to atmosphere the reactor will not suffer any catastrophic failures or radiation release.
A core meltdown type of event simply cannot happen; in the worst case the core will remain several hundred
degrees C below TRISO microsphere failure temperatures.

6. Plant Safety and Operational Performance
Most nuclear plant control systems today rely on operators to determine the correct course of action in complex
circumstances; nearly all reactor accidents and failures have been the result of incorrect operator techniques.
This is not practical for remote locations, and the StarCore Reactor Plant will be fully automatized using onsite hardware and software.
StarCore’s remote control technology will provide full-time monitoring and the ability to shut down the plantfrom StarCore Central or regional administrative centres by satellite links (2xGEO and 1xLEO) and reduce
on-site personal to only maintenance workers. This design for fully automated operation with ‘remotemonitoring and intervention (shutdown)’ is the StarCore Automated Reactor System (STARS) and has been
the subject of an independent review by the former Atomic Energy Canada Limited (AECL) and the Canadian
National Laboratories (CNL).

7. Instrumentation and Control Systems
StarCore owns the Intellectual Property to a modern fully automated control system design (the STARS
HyperVector Control System) previously used in many safety-critical aerospace systems. There are many
benefits that this control system technology brings, including automatic failure prediction for every system orcomponent in an arbitrarily complex application; alarms that uniquely identify any specific failures that haveoccurred or are predicted; controls that prevent wrong commands or actions ever being taken, and automatic
responses to arbitrarily complex failures.
The system is named after the manifold in n-dimensional state space that define the operational limits of thesystems and components; these are represented by n-vectors, or HyperVectors, defined as complex data in theimaginary plane. The states are defined for every component in the plant, recognize system operations, andpredict - in real time - any failures that may occur by calculating the state vector and time-to-state-operationalboundary for all components.

8. Plant Layout Arrangement
The plant has from two to six hardened silos at the base of the turbine hall; up to three turbine halls can beaccommodated in each plant. The main body of the plant is constructed of high-performance modules that arebonded into a single monocoque structure and is designed to be capable of resisting standard man-portableweapons such as RPGs. The plant is designed to meet a Beyond Design Basis Accident (such as an aircraft
or bomb) that destroys all above-ground facilities without causing any nuclear contamination.

153

9. Design and Licensing Status
StarCore has completed the initial design needed to develop Supplier Approval Procedures which resulted in
an Approved Supplier List and a Hardware Readiness Assessment. The StarCore Team includes direct-hired
StarCore personnel responsible for the overall management of the entire program, including all design
authority for commissioning, maintaining, operating, refuelling and decommissioning of the reactor plants.

10. Fuel Cycle Approach
The anticipated core lifetime is more than 5 years. At the end of every fifth year of operation, the graphite fuelprismatic blocks and spent fuel will be removed from the reactor pressure vessel and will be stored for 12months in an on-site underground fuel storage silo before being transported to a permanent repository site.
The spent fuel will be transported from the plant site to the repository site using a certified, existing fuel transfer
cask. Each transfer cask will hold six graphite fuel blocks, assembled in a Fuel Cartridge and which will be
replaced as one unit. StarCore will use a once-through LEU cycle initially, with plans to pursue R&D to moveto a fuel recycling capability depending on political agreements. We also plan to investigate TRU use in MOX
TRISO fuel to start recycling LWR waste in due course.

11. Waste Management and Disposal Plan
The plant will be decommissioned by StarCore at the end of its life, and all decommissioning cost is alreadybuilt into the StarCore Financial Plan. This work will include: removal of the reactor fuel and shipment to a
spent fuel storage facility; the spent fuel will be managed and disposed of; removal of the reactor vessel andshipment off site and disposed of; disposal of reactor components that cannot be reused; removal of all
equipment with radioactive components at the plant site and shipment to a disposal site; demolition of all above
ground facilities at the plant site and shipment of the materials off-site to a disposal facility; and entombmentof the below ground facilities at the plant site by backfilling them with concrete.

12. Development Milestones
209- 20132013-20172017-20202021-2026

154

Preliminary studies and initial pre-conceptual design.
Pre-conceptual design phase, technology validation and vendor contracts andqualification
Fund raising and PPA contracts
Projected deployment (start of construction to commissioning)JIMMY (JIMMY ENERGY SAS, France)
All content included in this section has been provided by, and is reproduced with permission of, JIMMY ENERGY SAS, France.

MAJOR TECHNICAL PARAMETERS
Parameter
Technology developer,
country of origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Value
JIMMY ENERGY SAS
FranceHigh temperature gas-cooled
ReactorHelium/ Graphite
10 to 20 / n.a.

reactorPrimary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
1.5 / 3.0

Core inlet/outlet coolant
temperature (oC)

300 / 700

Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)

UCO TRISO particles packed into pellets
19

Reactivity control
mechanism

B4C Control rods

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

Passive10 to 20150
5/5
15Zone 4
No refuelling, no on-site storage

Distinguishing features

Compact, low-weight design; simplicity
of manufacturing, flexibility of operation;
intrinsic passive safety

Design status

Detailed design

19.5
No refuelling110

1. Introduction
Jimmy is a high-temperature gas-cooled reactor with a thermal power of 10 to 20MW(t), designed to provide
competitive, low-carbon industrial process heat. The reactor is based on UCO TRISO particles, a graphite
moderator and a helium coolant (primary loop). A CO2 secondary loop allows to deliver heat up to 550°C toindustrial processes, selecting the secondary heat exchanger according to industrial needs. Valuing the existing
HTGR return of experience, the approach consists in delivering a simple, safe and robust heat generator. As aresult, the design relies mostly on well-tested solutions, as well on a passive safety demonstration, to optimize
the time- and cost-to-market. The first system will be available in 2026, with a first client site already identified.

2. Target Application
Jimmy targets the industrial heat market, where industrial suffer from both economic and environmentalpressure to reduce their use of fossil fuels, while no alternative solution available. Jimmy will allow to replaceindustrial gas-burners and provide competitive, low-carbon industrial process heat. The design and small
reactor footprint and limited capacity (up to 20MWt) allows to integrate into most industrial sites, with one orseveral generators in a row. The primary target is the French market, especially steam-consumers such as theAgro or Chemistry industry, with needs for steam up to 550°C. After massive deployment to process heatapplications in and outside of France, the main applications are district heating (offices, residentialbuildings…) and hydrogen production.

155

3. Design Philosophy
The main goal of the design is to provide a simple, safe and financially viable alternative to industrial heat gasburners. This goal is achieved by using HTGR well-tested solutions (especially regarding safety) and then
optimizing the design to optimize safety, reduce cost, and allow easy manufacturing, transportation andassembly. For instance, the use of a secondary CO2 loop allows to distinctly separate the nuclear perimeterfrom the industrial zone in terms of safety. Furthermore, the manufacturing and assembly aspects have beentaken into account since the beginning of the detailed design: design teams are in contact with presented
suppliers to screen commercially available components and integrate them in the design. It ensures that a
viable industrial scheme is available and to anticipate the orders of long-lead items, as well as to optimize
instruction time by safety authorities.

4. Main Design Features
Power Conversion
The Jimmy generator is designed to provide heat only, i.e., there is no energy conversion into electricity. A
primary exchanger allows to transfer heat from the primary helium-based loop to the secondary CO2-based
loop. The CO2-based loop pressure is 3.0 MPa and temperature is 600°C. A secondary exchanger is thenselected, according to each industrial site’s specific needs (fluid, temperature, pressure, etc.).

Reactor Core
The reactor core consists of 19 prismatic fuel assemblies inside a graphite reflector. Theses fuel assemblies aredesigned to be assembled in a factory and transportable inside existing fuel transportation packages.

Fuel Characteristics
The fuel are pellets of HALEU TRISO particles that are conformed to the AGR program envelope to reach the
expected burnup.

Reactivity Control
Reactivity control is made with 24 independent B4C control rods and B4C neutronic poisons that cross thecore or the reflector.

Reactor Pressure Vessel and Internals
The Jimmy reactor does not feature a pressure vessel but 19 different pressure tubes that constitute the core.
This whole core is then contained into a containment vessel. This geometry facilitates the transport andassembly of the reactor on its installation site.

Reactor Coolant System
The primary coolant is a helium pressurized circuit, whose circulation is forced on the cold branch by a helium
blower. The circuit divides and reunites before and after the pressure tubes of the core.

Secondary System
The secondary circuit connects to the primary circuit through a plate heat exchanger. It contains pressurized
carbon dioxide whose circulation is also forced.

Steam Generator
Jimmy’s generator does not directly produce steam, but its industrial client may produce steam based on theheat it gets from the secondary system through a secondary exchanger.

5. Safety Features
Engineered Safety System Approach and Configuration
Jimmy’s safety is mainly based on the leverage of three properties that the HTR can achieve: passive decay
heat removal, TRISO particles robustness and high reactivity coefficients. Based on these intrinsic properties,
most of the safety systems are around redundant and highly independent control rods.

Decay Heat Removal System / Reactor Cooling Philosophy
During normal operation, the reactor is cooled thanks to the industrial site that use the transferred heat for its
process.
During accidental operation, the graphite plays the role of a thermal absorber. The low power density of thereactor easily ensure that the fuel remains below the limit temperature that it can endure.

Spent Fuel Cooling Safety Approach / System
Calculations and modelling show that spent fuel from a Jimmy’s reactor cools passively and does not require
temporary on-site storage in a spent fuel pool.

156

Containment System
The design’s containment system is based on the “defence in depth” philosophy and includes a set of technicalmeasures aimed at preventing environmental exposure to radiation and radioactive substances from the reactor.
The set of barriers includes:- The coating layers of TRISO particles;- The graphite of the core
- The primary system- The containment of the building

Chemical Control
The main chemical control is a helium purification system that supplies a highly pure helium and monitor thehealth of the core

6. Plant Safety and Operational Performances
The goal of Jimmy is to ensure that large radioactivity release frequency is less than 10-7/reactor year.

7. Instrumentation and Control System
The I&C system is under design and does not contain any major disruption

8. Plant Layout Arrangement
The layout of Jimmy’s reactor is a building of w: 12m * l: 8m * h: 15 m. For security reasons, walls are thickto prevent aggression and only few systems go through the wall (mainly the RCCS and the secondary systemthat reaches the industrial line). All the access for maintenance and installation are on the roof to stay out ofan easy reach.

9. Testing Conducted for Design Verification and Validation
Jimmy’s philosophy is to use already tested technologies. Thus, most of the required tests are validation tests
on the behaviour on well modelled pieces. Jimmy does not plan to conduct a specific test program.

10. Design and Licensing Status
Interaction with Nuclear Safety Authority – Formal licensing process under way;
(Launch of licensing procedure with the first regulatory application (Dossier d’Options de Sûreté) in April
2020, allowing the French Safety Authority to start the instruction)
Site permit for FOAK plant – To be developed.
(Application to be submitted in June 2023 – First site for FOAK has already been identified, with seismic andother environmental studies underway)

11. Fuel Cycle Approach
In order to simplify the design and the operation, and to maximize safety, no refuelling will be done. After the
10-to-20-year life (depending on the nominal power chosen by the client), the entire vessel is extracted fromthe generator and replaced. Jimmy has a long-term strategy to foster fuel recycling between its systems.

12. Waste Management and Disposal Plan
By design, the Jimmy reactor does not release any waste during normal operation. The only possible source ofwaste would be the leaks from the helium loop, which is filtered and stored.
At the end of reactor lifetime, the reactor produces waste that can be collected and disposed of in accordancewith applicable regulation, relying on existing French waste management infrastructures with minimal
adaptations.

13. Development Milestones
2021

Done so far in 2022

Current goals

2023

Creation of the company
Seed fundraising allowing to finance Basic Design and first regulatory application
Basic Design completed
Peer review of Basic Design
Feasibility study for a first industrial client (Agro industry) completed
Launch of licensing procedure with the first regulatory application (Dossierd’Options de Sûreté) and instruction by the French safety authority (ASN)
Preliminary Design completed
Peer review of Preliminary DesignFirst orders placed for long-lead items
First design team completed (20+ engineers)
Detailed Design completed
Peer review of Detailed Design

157

2024
2025-2026

158

Second licensing application (Demande de Décret d’Autorisation de Création) andbeginning of the instruction by the French safety authority
Manufacturing Design completed
Peer review of Manufacturing Design
Orders placed for medium-lead items
Orders placed for all items
Equipment qualification
Communication among local population
Module pre-assembly (out of site)
On-site land development
Obtaining of licensing approval
On-site reactor assembly & tests
Fuel loading & tests
Commissioning

GTHTR300 (JAEA Consortium, Japan)
All content included in this section has been provided by, and is reproduced with permission of JAEA, Japan.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

JAEA, MHI, Toshiba/IHI,
Fuji Electric, KHI, NFI, Japan
Prismatic HTGR

Coolant/moderator

Helium / graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation<600 / 100~300

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
temperature (oC)
Fuel type/assembly array

7/7

Number of fuel assemblies
in the core
Fuel enrichment (%)(UO2)

Forced by gas turbine

587-633 / 850-950
UO2 TRISO ceramic coatedparticle
90
14

Refuelling Cycle (months)

48

Core Discharge Burnup(GWd/ton)
Reactivity control mechanism

120
Control rod insertion

Approach to safety systems

Active and passive

Design life (years)

60

Plant footprint (m2)~250 x 250 (4-reactor plant)

RPV height/diameter (m)23 / 8

RPV weight (metric ton)

~1 000

Seismic Design (SSE)>0.18g automatic shutdown

Fuel cycle
requirements / approachDistinguishing features

Uranium once through(initially)
Multiple applications of
power generation,cogeneration of hydrogen,process heat, steelmaking,desalination, district heating
Basic design

Design status

1. Introduction
The 300 MW(e) Gas Turbine High Temperature Reactor (GTHTR300) is a multi-purpose, inherently safe andsite-flexible small modular reactor (SMR) being developed by Japan Atomic Energy Agency (JAEA) for
commercialization in 2030s. As a Generation-IV technology, the GTHTR300 offers important advances overcurrent light water reactors. The coolant temperature is significantly higher in the range of 850-950oC. Suchhigh temperature capability as proven in the JAEA’s HTTR test reactor operation enables a wide range ofapplications. The design employs a direct-cycle helium gas turbine to simplify the plant by eliminating waterand steam systems while generating power with enhanced efficiency of 45-50%.

2. Target Applications
Typical applications include electric power generation, thermochemical hydrogen production, desalinationcogeneration using waste heat, and steel production. The reactor thermal power may be rated up to 600 MW(t).
The maximum hydrogen production per reactor is 120 t/d, enough to fuel about one million cars, 280-300
MW(e) electricity generation with additional seawater desalination cogeneration of 55 000 m3/d potable water
for about a quarter million of population, and annual production of 0.65 million tons of steel. All these are159

produced without CO2 emission.

3. Design Philosophy
The overall goal of the GTHTR300 design and development is to provide a family of system options capableof producing competitive electricity, hydrogen, desalination, other products, and yet deployable in the nearterm. The development of the multiple systems simultaneously does not necessarily suggest having investment
and risk multiplied. Rather, the development requirement is minimized by pursuing system simplicity,
economic competitiveness and originality, namely the SECO design philosophy.

4. Main Design Features
Power Conversion
The reactor system combines a high temperature gas-cooled reactor with direct-cycle gas turbine to generate
power while circulating the reactor coolant. The system consists of three functionally-oriented pressure vessel
units, housing the reactor core, the gas turbine, and the heat exchangers respectively. The multi-vessel systemfacilitates modular construction and independent maintenance access to the functional vessel units. The reactor
system is placed below grade in the reactor building. The pre-application basic design of the system wascompleted in 2003 by JAEA and domestic industrial partners Mitsubishi Heavy Industries, Fuji Electric,
Nuclear Fuel Industries and others. The reactor system design added cogeneration capabilities by adding an
IHX between the reactor and gas turbine that can accept the various roles of cogeneration while sharing
equipment designs with GTHTR300.

Reactor Primary Systemof the GTHTR300

Reactor Core
The reactor core consists of 90 fuel columns arranged in an annular ring, 73 and 48 inner and outer removablereflector columns, and 18 outer fixed reactor sectors. The effective annular core is about 3.6-5.5m in inner toouter diameter and 8m in height. The fuel column is a hexagonal graphite block similar to the HTTR fuel but
improved from it by an integral or sleeveless fuel rod for increased heat flux and by and enlarged coatedparticle buffer for improved fission product retention integrity under high burnup.

Fuel Characteristics
The fuel design is coated fuel particle of less than 1 mm in diameter. Each particle consists of a UO2 kernel
coated by four layers of low and high density pyro-carbon and silicon carbide. The all ceramic particle fuel isheat resistant up to 1600oC. Approximately 10 000 particles are packaged into a compact of the size of a thumb.
The compacts are then assembled into graphite-clad fuel rods. The fuel rods are inserted into the bore holes ofa hexagonal graphite fuel block of about 1 m long and 41 cm across, where the annulus formed between thefuel rod and the bore hole provides coolant flow channels. The fuel blocks are loaded into the reactor core.
The more fuel blocks are placed in the core, the higher the power output of the reactor.

Fuel Handling System
The fuel handling system consists of fuel loading machines, door valves, a control rod exchange machine, anda transport carriage. The fuel loading machine is used to remove fuel blocks from core and load fuel blocks tocore and spent fuel storage facility. The door valves are devised at the interface between fuel loading machineand spent fuel storage facility to maintain airtightness and radiation shielding. The control rod exchange
machine is installed for removal of control rods from reactor and loading of used control rods to maintenancepit. The transport carriage is used for the transportation of the fuel loading and control rod exchange machines.

Reactivity Control
Reactivity control system consists of control rods, control rod drive mechanisms and reserve shut down
systems. The system is used to adjust control rod position for reactivity control as well as shut down reactor in160

case of reactor scram. GTHTR300 has 30 pairs of control rod and reserve shut down systems. The control rods
and reserve shut down system channels are located in reflector blocks on inner and outer rings of fuel region.

Reactor Pressure Vessel and Internals
The reactor core consists of graphite hexagonal blocks, one-third of which are fuel blocks arranged in anannular region while the other two-thirds are reflector blocks arranged inside and outside of the fuel region.
Each fuel block has 57 coolant holes with fuel rods forming annular-shaped coolant channels. A permanent
reflector which surrounds side replaceable reflectors contains coolant channels. The helium coolant from thereactor inlet is introduced to the channel. A core barrel is installed between reactor internals and reactor
pressure vessel to support internal structure laterally.

Reactor Coolant System
Helium is heated in the reactor core to the cycle top temperature at high pressure. It then enters the turbine for
expansion to convert thermal energy into shaft power needed by the turbine to drive the compressor and electric
generator on single shaft. The turbine exhaust helium enters the recuperator, wherein its residual heat isrecovered in high effectiveness to preheat coolant to the reactor..

5. Safety Features
The reactor delivers fully inherent safety due to three enabling design features:- Ceramic coated particle fuel maintains containment integrity under a temperature limit of 1600oC.
- Reactor helium coolant is chemically inert and thus absent of explosive gas generation or phase change.
- Graphite-moderated reactor core provides negative reactivity coefficient, low-power density, and highthermal conductivity.
As a result of these features, the decay heat of the reactor core can be removed by natural draft air coolingfrom outside of the reactor vessel for a period of days or months without reliance on any equipment or operator
action even in such severe accident cases as loss of coolant or station blackout, while the fuel temperature willremain below the fuel design limit.

Engineered Safety System Approach and Configuration
An engineered safety system in the GTHTR300 consists of a reactor cavity cooling system and confinement.
GTHTR300 safety design is based on philosophy of maintaining safety functions relying on inherent andpassive safety features. Accordingly, the system is designed not to rely on active components or operator
actions in principle.

Decay Heat Removal System / Reactor Cooling Philosophy
GTHTR300 design removes the core decay and residual heat from the outside surface of the reactor vessel by
the natural convection and radiation, and to transfer it to an ultimate heat sink in the operational states and in
the accident conditions so that the design limits for fuel, the reactor coolant pressure boundary and structures
important to safety are not exceeded.

Spent Fuel Cooling Safety Approach / System
The fuel cycle is of once-through-then-out design allowing for 1460 days (4 years) of in-core fuel residencewith 120 GWd/ton average burnup. The spent fuel is stored in dry cooling facility with expenses included infuel cycle cost estimate. Fuel reprocessing is technically feasible and preliminary evaluation for this option is
being conducted.

Containment System
GTHTR300 employs a vented confinement rather than a conventional high-pressure, airtight containment usedin LWRs. The confinement is designed to meet requirements designated in safety design. The confinement isdesigned to release helium coolant blown out from primary system in case of depressurized loss-of-forced
cooling accidents. Dampers are devised and recloses when the pressure between outside and inside ofconfinement equalizes. The design leak rate is set to 20%/day to limit the amount of air ingress to reactor core.

Chemical Control
The helium purification systems are installed in the primary and secondary cooling systems in order to reduce
the quantity of chemical impurities such as hydrogen, carbon monoxide, water vapor, carbon dioxide, methane,
oxygen, and nitrogen. The primary helium purification system is mainly composed of a pre-charcoal trap, an
inlet heater, two copper oxide fixed beds, coolers, two molecular sieve traps, two cold charcoal traps andhelium compressors.

6. Plant Safety and Operational Performance
The ability to follow variable power and heat loads is simulated as shown in the following figure. The
simulation shows the plant response to an electric demand increase of 5%/min with corresponding reductionin heat rate, which is the maximum required for ramp load follow. The reactor remains at 100% power at all
times. Starting from a base cogeneration ratio of 203 MW(e) electricity and 170 MW(t) net heat, the turbine
power generation is increased to follow the increasing electric demand by increasing primary coolant inventory
161

1000

Reactor outlet temperature

40
20

950

0

900

4

800

3

Turbine inlet

600

2

400

Cycle pressure ratio

200

1

0

0

0

1

2

3

4

5

6

Elapsed time (min)

7

8

300

200

200

Electricity generation rate

100

100

IHX heat rate

0

0

20

150

15

100

CV4

50

10

0

5
0

IV1
0

1

2

3

4

5

6

Elapsed time (min)7

CV4 flow rate (kg/s)

1000

IV1 flow rate (kg/s)

1050

Reactor power

300

Heat rate (MWt)80
60

Cycle pressure ratio

Temperature (o C)

1100

100

Temperature (o C)

Reactor power (%)

120

Electricity rate (MWe)with opening of inventory valve IV1. The IHX heat rate is lowered by lowering secondary loop flow of the
IHX. The power sent out to external grid increases to 276 MW(e) in as little as a few minutes. The pressure in
the reactor increases to 7 MPa from 5 MPa.

8

7. Instrumentation and Control System
The instrumentation and control system consists of reactor and process instrumentations, control systems,
safety protection system, and engineered safety features actuating systems. The six (6) fundamental controls
are for turbine bypass, inventory, reactor outlet temperature, turbine inlet temperature, process heat supply rateand IHX differential pressure controls. They are combined in the basic plant control to provide controllability
for a variety of transients including loss-of-load and electric load following as shown above.

8. Plant Layout Arrangement
The reactor building is a below-grade, steel concrete structure housing 4 units of reactor systems consists ofsubsystems including reactor module, gas turbine module and heat exchanger modules.

9. Testing Conducted for Design Verification and Validation
The test results using the HTTR will be utilized for the development of GTHTR300. The test items cover fuelperformance and radionuclide transport, core physics, reactor thermal hydraulics and plant dynamics, andreactor operations, maintenance, control, etc. The results of the system performance analysis showed that thereactor could be continuously operated with the above variable load conditions. However, an actual
demonstration test is warranted for performance confirmation.

10. Design and Licensing Status
The design is developed at pre-licensing basic design stage. The design and development are planned to beconcluded to prepare for the demonstration plant operation in 2040s.

11. Fuel Cycle Approach
The design is applicable to fuel cycle options including UO2, MOX, and Pu-burning.

12. Waste Management and Disposal Plan
The design is applicable to options of direct disposal or reprocessing for spent fuel.

13. Development Milestones
2003
2005
2015
2020s
2040s

162

Basic design of GTHTR300 completed
Design for cogeneration plant GTHTR300C
Basic design for HTTR-connected gas turbine and H2 plant (HTTR-GT/H2) for systemdemonstration
HTTR-H2 test plant construction and operation (planned)
Operation of demonstration plant (TBD)

GT-MHR (JSC “Afrikantov OKBM”, RussianFederation)
All content included in this section has been provided by, and is reproduced with permission of, OKBM Afrikantov, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type

JSC “Afrikantov OKBM”,
Russian Federation
Modular Helium Reactor

Coolant/moderator

Helium /graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

600 / 288

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

7.2 / -

Number of fuel assemblies in
the core
Fuel enrichment (%)

Forced circulation

490 / 850
Coated particle fuel incompacts, hexagonal prism
graphite blocks of 0.36 m~1020
14-18% LEU or WPu

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

100-720 (depends on fuel
type)
25

Reactivity control mechanism

Control rod insertion

Approach to safety systems

Hybrid (active and passive)

Design life (years)

60

Plant footprint (m2)9110

RPV height/diameter (m)29 / 8.2

RPV weight (metric ton)

950

Seismic Design (SSE)

8 points (MSK 64)

Fuel cycle requirements /ApproachStandard LEU or WPu /No recycling; high fission
product retention
Inherent safety characteristics;
no core melt; high temperatureprocess heat capabilities; small
number of safety systems
Preliminary design completed;
key technologies are being
demonstrated

Distinguishing features

Design status

1. Introduction
The gas turbine modular helium reactor (GT-MHR) couples a HTGR with a Brayton power conversion cycle
to produce electricity at high efficiency. As the reactor unit can produce high coolant outlet temperatures, themodular helium reactor system can also efficiently produce hydrogen, e.g. by high temperature electrolysis orthermochemical water splitting.
The use of modular helium reactor units makes the system flexible and allows to use various power conversion
schemes: with gas-turbine cycle, steam-turbine cycle and with a circuit supplying high-temperature heat toindustrial applications. The modular high temperature gas-cooled reactor unit possesses inherent safety
features with safe passive removal of decay heat providing a high level of safety even in the case of total lossof primary coolant.
The modular helium reactor design proved unit modularity with a wide power range of a module (from 200 to
600 MW(t)) and NPP power variation as a function of module number. This provides good manoeuvringcharacteristics of the reactor plant (RP) for regional power sources.
163

2. Target Application
The GT-MHR can produce electricity at high efficiency (approximately 48%). As it can produce high coolant
outlet temperatures, the modular helium reactor system can also efficiently produce hydrogen, e.g. by hightemperature electrolysis or thermochemical water splitting.

3. Design Philosophy
The GT-MHR direct Brayton cycle power conversion system contains a gas turbine, an electric generator andgas compressors. The layout can be seen in the figure above. The GT-MHR gas turbine power conversion
system has been made possible by utilizing large, active magnetic bearings, compact, highly effective gas togas heat exchangers and high strength, high temperature steel alloy vessels. The use of the gas-turbine cycle
application in the primary circuit leads to a minimum number of reactor plant systems and components. The
GT-MHR safety design objective is to provide the capability to reject core decay heat relying only on passive(natural) means of heat transfer without the use of any active safety systems. The GT-MHR fuel form presents
formidable challenges to diversion of materials for weapon production, as either fresh or as spent fuel.

4. Main Design Features
Power Conversion System Flow Diagram
1. Reactor
2. Gas Turbine
3. Recuperator
4. Pre-Cooler
5. Low Pressure Compressor
6. Intercooler
7. High Pressure Compressor
8. Generator
9. Generator cooler
10. Bypass Valve11~15. SCS Components
16. Surface Cooler Of reactor
17. Air Ducts
18. Heat Exchanger with Heat pipes
The Brayton power conversion with direct gas turbine contains a gas turbine, an electric generator and gas
compressors. The GT-MHR gas turbine power conversion system has been made possible by utilizing large,
active magnetic bearings, compact, highly effective gas to gas heat exchangers and high strength, hightemperature steel alloy vessels.

Reactor Core and Fuel Characteristics
Coated particle fuel is used. The fuel kernel (U or Pu oxide) is coated by a first porous layer of pyrocarbon,
followed by a dense layer of pyrocarbon, a silicon carbide layer and an outer dense layer of pyrocarbon.
Thousands of coated particles and graphite matrix material are made into a fuel compact with thousands ofcompacts inserted into the fuel channels of the hexagonal prismatic graphite blocks or fuel assemblies. The
coated particles will contain almost all fission products with temperatures up to 1600oC. About 1 billion fuelparticles of the same type were manufactured and tested in Russia. The standard fuel cycle for the commercial
GT–MHR utilizes low enriched uranium (LEU) but alternative cycles including the disposition of plutonium
were also studied in detail.

Reactivity Control
Two independent reactivity control systems based on different operation principles are used to execute reactor
emergency shutdown and maintenance in a sub-critical state. These systems are: 1) Electromechanicalreactivity control system based on control rods moving in the reactor core channels and in the inner and outerreflectors; 2) Reserve Shutdown System (RSS) based on spherical absorbing elements that fill-in channels in
the fuel assemblies stacked over the whole height of a fuel assembly. Control rods with boron carbide
absorbing elements located in the reflector are used during normal operation and hot shutdown, and rods
located in the core are used for scram.

Reactor Pressure Vessel and Internals
The reactor pressure vessel, made of chromium-molybdenum steel, is 29 m in height with an outer diameter(across flanges) of 8.2 m. Prerequisites and conditions excluding brittle fracturing of the reactor vessel
include keeping the fast neutron fluence on the reactor vessel and the vessel temperature below theallowable limits. In-vessel structures, namely, prismatic fuel blocks, reflectors, and core support structure

are made of graphite, and metallic structures are made of chromium-nickel alloy. Service life of the reactor
vessel and internals is 60 years.
1645. Safety Features
Safety objectives for the GT-MHR are achieved, first, by relying on the inherent safety features incorporated
in the plant design. The design features, which determine the inherent safety and ensure thermal, neutronic,
chemical and structural stability of the reactor unit, are the following:
Using of helium coolant, which has some specific properties. During plant operation, helium is not affected byphase transformations, does not dissociate, is not activated and has good heat transfer properties. Helium ischemically inert, does not react with fuel, moderator and structural materials. There are no helium reactivity
effects;
Core and reflector structural material is high-density reactor graphite with substantial heat capacity and heat
conductivity and sufficient mechanical strength that ensures core configuration preservation under any
accident;
Nuclear fuel in the form of coated fuel particles with multilayer ceramic coatings, which retain integrity andeffectively contain fission products under high fuel burnup and high temperatures;
The temperature and power reactivity coefficients are negative, what provides the reactor safety in any design
and accident conditions.
Safety is ensured by application of passive principles of system actuation. The decay and accumulated heat isremoved from the core through reactor vessel to reactor cavity cooling system and then to atmosphere by
natural physical processes of heat conductivity, radiation, convection without excess of fuel safe operation
limits including LOCA, in case of all active circulation systems and power sources failure.

(a) Engineered Safety System Approach and Configuration
In addition to the inherent (self-protection) features of the reactor, the GT-MHR plant incorporates safetysystems based on the following principles: 1) Simplicity of both system operation algorithm and design; 2)
Use of natural processes for safety system operation under accident conditions; 3) Redundancy, physicalseparation and independence of system channels; 4) Stability to the internal and external impacts andmalfunctions caused by accident conditions; 5) Continuous or periodical diagnostics of system conditions; 6)Conservative approach used in the design, applied to the list of initiating events, to accident scenarios, and forthe selection of the definitive parameters and design margins.

(b) Reactor Cooling Philosophy

High heat storage capacity of the reactor core and high acceptable temperatures of the fuel and graphite allow
passive shutdown cooling of the reactor during accidents, including LOCA (heat removal from the reactor
vessel by radiation, conduction and convection), while maintaining the fuel and core temperatures within theallowable limits. The GT-MHR design provides for no dedicated active safety systems. Active systems ofnormal operation, such as the power conversion unit and the shutdown cooling system are used for safety
purposes. These systems remove heat under abnormal operation conditions, during design basis accidents(DBA) and in beyond design basis accidents (BDBA). Emergency heat removal can also be carried out by thereactor cavity cooling system (RCCS). Heat from the reactor core is removed through the reactor vessel to the
RCCS surface cooler, the heat tubes and then to the atmospheric air due to natural processes of heat conduction,
radiation and convection. Water and air in the RCCS channels circulate driven by natural convection.

(c) Containment Function

Passive localization or radioactivity is provided by the containment designed for the retention of helium-air
fluid during accidents with primary circuit depressurization. The containment is also designed for the externalloads, which may apply to seismic impacts, aircraft crash, air shock wave, etc. Activity release from thecontainment into the environment is determined by the containment leakage level, which is about 1 % of thevolume per day at an emergency pressure of 0.5 MPa.

6. Plant Safety and Operational Performances
All safety systems are designed with two channels. Fulfilment of the regulatory requirements on safety, provenby a compliance with both deterministic and probabilistic criteria, is secured by an exclusion of the active
elements in a channel or by applying the required redundancy of such active elements inside a channel, as well
as via the use of the normal operation systems to prevent design basis accidents.

7. Instrumentation and Control Systems
The GT-MHR NPP control and support safety systems (CSS) are intended to actuate the equipment,
mechanisms and valves, localizing and support safety systems in the pre-accidental conditions and in accidents;
to monitor their operation; and to generate control commands for the equipment of normal operation systems
used in safety provision algorithms. The CSS are based on the principles of redundancy, physical andfunctional separation, and safe failure. The CSS sets are physically separated so that internal (fire, etc.) or
external (aircraft crash, etc.) impacts do not lead to a control system failure to perform the required functions.
The CSS provide automated and remote control of the equipment of safety systems from the independent mainand standby control rooms. Principal technical features are selected using the concept of a safe failure blackouts, short-circuits, or phase breaks initiate emergency signals in the channels or safety actions directly.

165

8. Plant Layout Arrangement
The plant layout is shown on the right.

9. Design and Licensing Status
Reactor plant preliminary design and demonstrationof key technologies for Pu-fuelled option
completed.

10. Fuel Cycle Approach
The GT-MHR fuel cycle approach is a once throughmode without reprocessing. Fuel handling
operations are performed using the protective
containers to avoid fuel assembly damage andradioactive product release. Appropriately shieldedcontainers are provided to protect the personnel
against radiation impacts during dismantling of thereactor unit components at fuel reloading. These
measures are also applied at spent fuelmanagement. Spent fuel shows good proliferation
resistance characteristics, producing less materials
of proliferation concern (total plutonium and 239Pu)
per unit of energy produced.

11. Waste Management and Disposal Plan
Facilities for long-term storage of spent nuclear fuel (SNF) and solid/solidified radioactive waste (RW) are
included in the complex of a GT-MHR commercial 4-unit NPP. The capacity of the designed SNF storage isdetermined from the condition of capability to store fuel unloaded from the NPP for 10 years. The estimatedtotal construction volume of the SNF reception and storage compartments is around 150 000 m3. The capacityof solid/solidified RW storage facility is designed to provide storage of waste generated during the 10-year
period of NPP operation. After 10 years of storage at the NPP site, SNF and RW are to be removed for final
underground disposal. Radiochemical SNF reprocessing is considered as an option for future only.

12. Development Milestones
1993
1994
1996
1997
1998
1999
2000
2002
2002
2003
2014
Since 2014

166

Minatom / General Atomics MOU on joint GT-MHR development for commercial units
Russia proposes to build GT-MHR at Seversk to burn Russian WPu
Framatome& Fuji Electric join the GT-MHR program
Conceptual design completed
GT-MHR becomes an option within the US/RF Pu disposition strategy
Conceptual design review by international group of experts
Work started on preliminary design
Project review by Minatom of Russia and US DOE experts
Reactor plant preliminary design completed
Begin demonstration of key technologies
Completion of demonstration of key technologies for Pu-fuelled core
Use of principal reactor unit design features as a basis for MHR-T design (U-fuelled
option)MHR-T Reactor (JSC “Afrikantov OKBM”,
Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of JSC “Afrikantov OKBM”, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type
Coolant/moderator

JSC “Afrikantov OKBM”,
Russian Federation
Modular helium hightemperature reactor
Helium/graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

4 × 600 / 4 × 205.5

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

7.5 / –

Forced circulation

578 / 950

Number of fuel blocks

Coated particle fuel incompacts, hexagonal prism
graphite block of 0.36m~ 1 020

Fuel enrichment (%)< 20Core Discharge Burnup(GWd/ton)
Fuel cycle (months)

125

Reactivity control mechanism

Control rods

Approach to safety systems

Hybrid (active and passive)

Design life (years)

60

RPV height/diameter (m)32.8 / 6.9

RPV weight (metric ton)

950

Seismic design

8 points (MSK 64)

Fuel cycle requirements /ApproachStandard LEU /No recycling; high fission
product retention
Multi-module HTGR dedicatedto hydrogen production / hightemperature process heatapplication.
Conceptual design

Distinguishing features

Design status30

1. Introduction
The MHR-T reactor/hydrogen production complex makes use of the basic GT-MHR reactor unit design as thebasis for a multi-module plant for hydrogen production. The hydrogen production through the steam methane
reforming process or high-temperature solid oxide electrochemical process is performed by coupling the plantwith the modular helium reactor(s). The use of modular helium reactor units makes the system flexible andallows the possibility to use various power unit schemes: with gas-turbine cycle (GT-MHR design), steamturbine cycle and with the circuit supplying high-temperature heat to industrial applications (this design). The
modular high temperature gas-cooled reactor unit possess salient safety features with passive decay heatremoval providing high level of safety even in case of total loss of primary coolant.

2. Target Application
The most perspective technologies for Russia are hydrogen production through the steam methane reforming
process or high-temperature solid oxide electro-chemical process coupled with a modular helium reactor called
MHR-T. The chemical-technological sector with steam methane reforming is considered as an option for shortterm perspective.

167

3. Design Philosophy
The MHR-T complex includes the chemical-technological sector (hydrogen production sector) and theinfrastructure supporting its operation. The chemical-technological sector includes hydrogen production
process lines, as well as systems and facilities supporting their operation.
The following processes are considered as the basic processes for the chemical-technological sector: (i) steam
methane reforming; and (ii) high-temperature solid oxide electrochemical process of hydrogen production
from water. Heat shall be transferred directly from primary coolant to chemical-technological sector medium
in a high-temperature heat exchanger. The key component of chemical-technological sector medium
circulating through the high-temperature heat exchanger is water steam. The high-temperature electrolysis
option allows the consideration of two- and three-circuit plant configurations. The technical concept is basedon:- Modular helium-cooled reactors with typical high level of inherent safety;- Fuel cycle based on uranium dioxide in the form of multi-layer coated particles, high burnup and burial offuel blocks unloaded from the reactor without any additional processing;- Electromagnetic bearings operating almost without friction and applied in various technical areas;- Highly efficient high-temperature compact heat exchangers, strong vessels made of heat resistant steel.
The thermal energy generated in the reactor is converted to chemical energy in a thermal conversion unit
(TCU) where, in the MHR-T option with methane reforming, the initial steam-gas mixture is converted tohydrogen-enriched converted gas (mixture of water steam, СО, Н2, СО2, and СН4) in the course of athermochemical reaction.

4. Main Design Features
Power Conversion System
A power conversion unit isintegrated in a single vessel andincludes a vertical turbomachine,
highly efficient plate heatexchanger, and coolers. A hightemperature heat exchanger (IHX)
for the MHR-T option withmethane steam reforming is an
integral part of the thermal
conversion unit and is partitioned
as a three-stage heat exchanger.
Arrangement
of
the
heat
exchanger sections along the
primary coolant flow is parallel,and downstream of the coolant in
the chemical-technological sector(CTS) is sequential. Each section
is designed as a separate heatexchanger consisting of several
modules.
Power Conversion UnitHigh Temperature Heat Exchanger Section
The material of the heat exchange
surface of the module is a heatresistant alloy. The turbomachine includes a generator and turbo-compressor mounted on a single shaft onelectromagnetic suspension. The gas turbine cycle of power conversion with the helium turbomachine, heatexchanger and intercooler provide a thermal efficiency of 48%.

Reactor Core and Fuel Characteristics
Coated particle fuel is used. The fuel kernel (U oxide) is coated by a first porous layer of pyrocarbon, followed
by a dense layer of pyrocarbon, a silicon carbide layer and an outer dense layer of pyrocarbon. Thousands ofcoated particles and graphite matrix material are made into a fuel compact with thousands of compacts inserted
into the fuel channels of the Hexagonal Prism graphite blocks or fuel assemblies.
The coated particles will contain almost all fission products with temperatures up to 1600oC. About 1 billion
fuel particles of the same type were manufactured and tested in Russia. The standard fuel cycle is to utilize
low enriched uranium (LEU) in a once through mode. The MHR-T show good proliferation resistancecharacteristics. It produces less total plutonium and 239Pu (materials of proliferation concern) per unit of energy
produced. The fuel form presents formidable challenges to diversion of materials for weapons production, as
either fresh or as spent fuel.

Reactor Coolant System
Working media in circulation circuits are helium of the primary circuit and steam-gas mixture (SGM) in the
CTS. The peculiarity of heat exchangers for the production of hydrogen by methane reforming process is the168transfer of heat from high temperature helium of the primary circuit to the chemically aggressive medium ofhydrogen production circuit.

Reactivity Control
Two independent reactivity control systems based on different operation principles are used to execute reactor
emergency shutdown and maintenance in a sub-critical state. These systems are: 1) Electromechanicalreactivity control system based on control rods moving in the reactor core channels and in the inner and outerreflectors; 2) Reserve shutdown system based on spherical absorbing elements that fill-in channels in the fuelassembly stack over the whole height of a fuel assembly. Control rods with boron carbide absorbing elements
located in the reflector used during normal operation and hot shutdown, and rods located in the core used forscram.

Reactor Pressure Vessel and Internals
Reactor Pressure Vessel made of chromium-molybdenum steel is 29 m height with outer diameter (acrossflanges) 8.2 m. Prerequisites and conditions excluding brittle fracturing of the reactor vessel include
keeping the fast neutron fluence on the reactor vessel and the vessel temperature below the allowable
limits. In-vessel structures, namely, prismatic fuel blocks, reflectors, and core support structure are made ofgraphite, and metallic structures are made of chromium-nickel alloy. Service life of the reactor vessel andinternals is 60 years.

5. Safety Features
The safety features of the MHR-T reactor are the same as for the GT-MHR. Safety objectives for the MHR-T
are achieved, first of all, by relying on the inherent safety features incorporated in the plant design. The design
features are as follows:- Using helium as the coolant. During operation, helium is not affected by phase transformations. It does notdissociate and has good heat transfer properties. Helium is chemically inert. It does not react with fuel,moderator and structural materials. There are no helium reactivity effects;- Core and reflector structural material is high-density reactor graphite with substantial heat capacity and heat
conductivity and sufficient mechanical strength that maintain core configuration integrity;- Nuclear fuel in the form of coated fuel particles with multilayer ceramic coatings, which retain integrity andeffectively contain fission products under high fuel burnup and high temperatures;- The temperature and power reactivity coefficients are negative that provides the reactor safety in any design
and accident conditions.
Safety is ensured by application of passive principles of system actuation. The decay and accumulated heat isremoved from the core through reactor vessel to reactor cavity cooling system and then to atmosphere by
natural physical processes of heat conductivity, radiation, convection In case of LOCA with failures of allactive circulation systems and power sources, operation limits of the fuel are not exceeded.

Engineered Safety System Approach and Configuration
Special considerations are devoted to external impacts from the hydrogen production sector. In addition to theinherent (self-protection) features of the reactor, the MHR-T plant incorporates safety systems based on thefollowing principles: 1) Simplicity of both system operation algorithm and design; 2) Usage of naturalprocesses for safety system operation under accident conditions; 3) Redundancy, physical separation andindependence of system channels; 4) Stability to the internal and external impacts and malfunctions caused by
accident conditions; 5) Continuous or periodical diagnostics of system conditions; 6) Conservative approach
used in the design, applied to the list of initiating events, to accident scenarios, and for the selection of thedefinitive parameters and design margins.

Reactor Cooling Philosophy
High heat storage capacity of the reactor core and high acceptable temperatures of the fuel and graphite allow
passive shutdown cooling of the reactor in accidents, including LOCA (heat removal from the reactor vesselby radiation, conduction and convection), while maintaining the fuel and core temperatures within theallowable limits. The MHR-T follows the GT-MHR design principles with no dedicated active safety systems(active systems of normal operation are used for safety purposes) and with emergency heat removal also
possible by the reactor cavity cooling system (see GT-MHR for more details).

Containment Function
The approach is the same as for the GT-MHR with passive localization or radioactivity provided by thecontainment as well as external loads (see GT-MHR for more details).

6. Plant Safety and Operational Performances
All safety systems are designed with two channels. Fulfilment of the regulatory requirements on safety, provenby a compliance with both deterministic and probabilistic criteria, is secured by an exclusion of the active
elements in a channel or by applying the required redundancy of such active elements inside a channel, as well
as via the use of the normal operation systems to prevent design basis accidents.

1697. Instrumentation and Control Systems
The MHR-T NPP control and support safety systems (CSS) are intended to actuate the equipment, mechanisms
and valves, localizing and support safety systems in the pre-accidental conditions and in accidents; to monitor
their operation; and to generate control commands for the equipment of normal operation systems used insafety provision algorithms. The CSS are based on the principles of redundancy, physical and functional
separation, and safe failure.

8. Plant Arrangement
The main components of each NPP module are arranged in isolated premises of the underground containmentof the NPP main building. The chemical-technological sector equipment is arranged outside the containmentof the NPP main building. The MHR-T energy-technological complex is designed for a specific site on thebasis of design solutions selected with account of climatic conditions typical of central Russia and special
external impacts such as seismicity, aircraft crash, air shock wave. The interfaces between the four-module
NPP and the chemical-technological sector must be designed to except faults that could cause failure of morethan one MHR-T module. The main reactor equipment is arranged in a vertical vessel located in a separate
cavity parallel to the power conversion unit and high-temperature heat exchanger vessels.

9. Design and Licensing Status
Feasibility study of plant application for large-scale hydrogen production completed. Currently, safety issues
are the major area of R&D with the emphasis on mutual influence of nuclear and hydrogen production
components of the facility.

10. Fuel Cycle Approach
The MHR-T fuel cycle approach is a once through mode without reprocessing. Appropriately shieldedcontainers are provided to protect the personnel against radiation impacts during dismantling of the reactor
unit components at fuel reloading. These measures are also applied at spent fuel management.

11. Waste Management and Disposal Plan
Facilities for long-term storage of spent nuclear fuel (SNF) and solid/solidified radioactive waste (RW) are
included in the complex of an MHR-T commercial 4-unit NPP. The capacity of the designed SNF storage isdetermined from the condition of capability to store fuel unloaded from the NPP for 10 years. The estimatedtotal volume of the SNF reception and storage compartments is around 150 000 m3. The capacity ofsolid/solidified RW storage facility is designed to provide storage of waste generated during the 10-year periodof NPP operation. After 10 years of storage at the NPP site, SNF and RW are to be removed for final
underground disposal.

12. Development Milestones
2001
2005
2007
2017
2020

170

Pre-conceptual proposal
Conceptual design completed
Elaboration of technical requirements
Feasibility study of plant application for large-scale hydrogen production
Development of basic design solutions for chemical-technological part of the
facilityMHR-100 (JSC “Afrikantov OKBM”,
Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of JSC “Afrikantov OKBM”, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

JSC “Afrikantov OKBM”,
Russian Federation
Modular helium reactor

Coolant/moderator

Helium/graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

215 / 25 – 87 (depends on
configuration)
Forced circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array4 – 5 / depends on configuration

Fuel enrichment (%)

490 – 553 / 795 – 950(depends on configuration)
Hexagonal prism graphite blocks
with coated particle fuelAbout 1600 blocks with more than
600 fuel compacts in each block< 20%

Reactivity control mechanism

Control rod insertion

Approach to safety systems

Hybrid (active and passive)

Design life (years)

60

Plant footprint (m2)

Depends on configuration

RPV height/diameter (m)Similar to 1000 MW(e) LWR

RPV weight (metric ton)

Similar to 1000 MW(e) LWR

Seismic Design (SSE)

8 points (MSK 64)

Fuel cycle requirements /ApproachDistinguishing features

Once through U; Pu and Th cycle
also possible
A multipurpose reactor forcogenerations of electricity, heatand hydrogen; high-temperature
heat supply to oil refinery plant
Conceptual design

Number of fuel assemblies

MHR-100 GT layout

Design status

1. Introduction
The designs are based on the global experience in the development of experimental HTGR plants. Russia hasmore than 40 years of experience in the development of HTGR plants of various power (from 100 to 1000MW) and for various purposes. It has established the experimental facilities for the R&D work, the fuel element
and material fabrication technology, including the fabrication and mastering of pilot equipment, and various
activities in hydrogen generation technology. Today, conventional power stations, with electric capacity ~300
MW(t), are deployed all over the territory of Russia. These are adapted to regional systems and provide theelectric power needs. They consist mainly of cogeneration plants producing about 40% of electric power and
85% of heat generation. Analysis shows that SMRs with HTGR have therefore good prospects to add to or
replace these regional generation. Innovative nuclear power systems to be implemented on this basis are
therefore considered as an important area of the nuclear power industry development up to the middle of the
century. Based on study of the power market development and demands, pre-conceptual work is performed
for commercial MHR100 with modular helium reactor and several power conversion layouts for various
power-industrial applications. The following options of MHR100 for industrial applications were studied:- Electricity and district heat productions by core thermal power conversion to electric one in direct gasturbine Brayton cycle – MHR-100 GT;- Electricity and hydrogen generations by high-temperature steam electrolysis – MHR-100 SE;- Hydrogen generation by steam methane reforming method MHR-100 SMR;
- High-temperature heat supply to oil refinery plant – MHR-100 OR.

171

2. Target Application
The MHR-100 is intended for regional power generation and heat production in the Russian Federation. A
single reactor unit design can be implemented in various plant configurations.
Major Technical Parameters of MHR-100 GT
Parameters
Reactor heat capacity (MW)Net power generation efficiency (%)
Helium temperature at reactor inlet/outlet (°C)
Low-pressure helium temperature at recuperator inlet (°C)
Helium flow rate through the reactor (kg/s)
Helium bypass flow rate from HPC outlet to recuperator outlet at
high-pressure side (kg/s)
Helium pressure at reactor inlet (MPa)Expansion ratio in turbine
Generator/TC rotation speed (rpm)
PCU cooling water flow rate (kg/s)
Delivery water temperature at inlet/outlet (°C)

Power Mode215
46.1
558 / 850
583
139.1
–

Cogeneration Mode215
25.4
490 / 795
595
134
32.2

4.91
2.09
3000 / 9000
804
–

4.93
1.77
3000 / 9000
480
70 / 145

3. Design Philosophy
The reactor power and its design are universal for all the different power and process heat options with only
the coolant parameters that are different. The reactor unit power level (215 MW(t)) was selected according to:(i) the regional power industry and district heat supply needs; (ii) the manufacture needs in high- and mediumtemperature heat supply for technological processes; and (iii) process capabilities of national enterprises infabrication of RP main components including vessels.

4. Specific Design Features
Power Conversion Unit
A power conversion unit is integrated in a single vessel and includes a vertical turbomachine, highly efficientheat exchanger, and coolers. The turbomachine includes a generator and turbo-compressor mounted on a single
shaft on electromagnetic suspension. Gas turbine cycle of power conversion with the helium turbomachine,
heat exchanger and intercooler provide thermal efficiency at 48%.

Reactor Core and Fuel Characteristics
Coated particle fuel is used. The fuel kernel (U oxide) is coated by a first porous layer of pyrocarbon, followed
by a dense layer of pyrocarbon, a silicon carbide layer and an outer dense layer of pyrocarbon. Thousands ofcoated particles and graphite matrix material are made into a fuel compact with hundreds of compacts inserted
into fuel channels of each hexagonal prism graphite block (0.2 m × 0.65 m height). The core is a cylindricalarrangement of vertical stacks of fuel blocks (fuel columns). The standard fuel cycle is to utilize low enricheduranium (LEU) in a once through mode.

Reactor Unit

172

Power Conversion Unit

Reactivity Control
The two (2) independent reactivity control systems are used to perform reactor emergency shutdown andmaintenance in a sub-critical state: (i) Electromechanical reactivity control system based on control rods in thechannels and in the inner and outer reflectors; (ii) Reserve shutdown system based on spherical absorbing
elements that fill-in channels in the fuel assembly stack over the whole height of a fuel assembly. Control rods
with boron carbide absorbing elements located in the reflector are used during normal operation and hot
shutdown, and rods located in the core are used for scram.

Reactor Pressure Vessel and Internals
The Reactor Pressure Vessel (RPV) made of chromium-molybdenum steel has dimension similar to that ofstandard VVER-1000. Prerequisites and conditions excluding brittle fracturing of the RPV include keeping thefast neutron fluence on and temperature of the RPV below the allowable limits. In-vessel structures, namely,
prismatic fuel blocks, reflectors, and core support structure are made of graphite. Metallic internals are madeof chromium-nickel alloy. Service life of the RPV and internals is 60 years.

5. Safety Features
Safety objectives for the MHR-100 are achieved, first of all, by relying on the inherent safety features. Thesedesign features ensure thermal, neutronic, chemical and structural stability of the reactor. Safety is ensured bypassive principles of system actuation. The decay and accumulated heat is removed from the core through RPV
to reactor cavity cooling system and then to atmosphere by natural heat conductivity, radiation, and convection.
In LOCA condition with failures of all active circulation systems and power sources, operation safety limit of
the fuel is not exceeded.

Engineered Safety System Approach and Configuration
In addition to the inherent features, the MHR-100 incorporates safety systems based on: (i) Simplicity of both
system operation algorithm and design; (ii) Natural processes for safety system operation under accident
conditions; (iii) Redundancy, physical separation and independence of systems; (iv) Stability to the internaland external impacts and malfunctions caused by accident conditions; (v) Continuous or periodical diagnostics
of system conditions; (vi) Conservative approach used in the design, applied to the list of initiating events, toaccident scenarios, and for the selection of the definitive parameters and design margins.

Decay Heat Removal / Reactor Cooling Philosophy
High heat storage capacity of the reactor core and high acceptable temperatures of the fuel and graphite allow
passive shutdown cooling of the reactor in accidents, including LOCA (heat removal from the reactor vesselby radiation, conduction and convection), while maintaining the fuel and core temperatures within theallowable limits The MHR-100 design provides for no dedicated active safety systems. Active systems ofnormal operation, such as the power conversion unit and the shutdown cooling system are used for safety
purposes. These systems remove heat under abnormal operation conditions, during design basis accidents(DBA) and in beyond DBA.

Containment Function
Passive localization of radioactivity is provided by the containment designed for the retention of helium-air
fluid during accidents with primary circuit depressurization. The containment is also designed for externalloads due to seismic impacts, aircraft crash, air shock wave, etc. Activity release from the containment into theenvironment is determined by the containment leakage level, which is about 1%vol/d day at 0.5 MPa.

6. Plant Safety and Operational Performances
All safety systems are designed with two channels. Fulfilment of the regulatory requirements on safety, provenby a compliance with both deterministic and probabilistic criteria, is secured by an exclusion of the active
elements in a channel or by applying the required redundancy of such active elements inside a channel, as well
as via the use of the normal operation systems to prevent design basis accidents.

7. Instrumentation and Control
Control and support safety systems are intended to actuate the equipment, mechanisms and valves, localizing
and support safety systems in the pre-accidental conditions and in accidents; to monitor their operation; and to
generate control commands for the equipment of normal operation systems used in safety provision algorithms.

8. Design Variants and Plant Arrangements Based on the Modular MHR-100
The modular reactor consists of the core with hexahedral prismatic fuel assemblies, uses helium as a coolant,
and has inherent self-protection. The technical concept of studied reactor plant MHR-100 is based on:- Modular high-temperature helium-cooled reactors with typical high level of inherent safety;- Fuel cycle with fuel in the form of multilayer UO2-based coated fuel particles, high burnup and possibility
to dispose the spent fuel blocks without additional reprocessing;- High-performance high-temperature compact heat exchangers, high-strength casings of heat-resistant steel;- Direct gas-turbine cycle with high-efficiency recuperation and intermediate coolant cooling;173

- Experience in high-efficiency gas turbines application in power engineering and transport;- Electromagnetic bearings used in power conversion system.
The coolant is circulated in the primary loops by the main gas circulator or by the power conversion unit (PCU)
turbomachine (TM) compressors. The MHR-100 option consists of power and process parts. The power partis unified to the maximum for all options and is a power unit consisting of a reactor unit with a thermal powerof 215 MW and a gas-turbine PCU for power generation and (or) heat-exchange units, depending on the
purpose. The process part of MHR-100 is either a process plant for hydrogen production or circuits for hightemperature heat supply to various technological processes, depending on the purpose.
The unified gas-turbine PCU is planned to be used in MHR-100 GT and MHR-100 SE options. Verticaloriented TM is the main feature of the PCU and consists of the turbo-compressor (TC) and generator withrotors, which have different rotation speed of 9000 rpm and 3000 rpm respectively. Electromagnetic bearings
are used as the main supports. The generator is located in air medium outside the helium circuit. The PCU precooler and intercooler are arranged around TC while the recuperator is located at the top of the vessel above
the hot duct axis. Waste heat from the primary circuit is removed in the PCU pre-cooler and intercooler by thecooling water system, then in dry fan cooling towers to atmospheric air.

MHR-100 SE

Heat exchange blocks are intended to transfer heat power from the reactor to the consumer of powertechnological applications. Depending on the working fluid, process type and probability of radioactivity
ingress to the process product and contamination of equipment with radioactive products, two- or three-circuit
RP configuration can be used. So, two circuit configurations are used in MHR-100 SE NPP for hydrogengeneration and in MHR-100 SMR for steam methane reforming. Water steam is the main component of process
fluid in these processes. The analysis shows that the effects of hydrogen-bearing products ingress in potential
accidents with depressurization of the steam generator (SG) or high-temperature heat exchanger (HX) arereliably checked by reactor control and protection systems.
MHR-100 OR-based power source for heat supply to petrochemical applications and oil refinery plants has
three-circuit thermal configuration. Heat from RP is transferred to the consumer via high-temperatureintermediate helium-helium HX (IHX) and intermediate helium circuit and then to network circuit ofpetrochemical applications. This decision restricts radioactivity release to the network circuit and provides
radiological purity of the process product and minimum contamination of the primary circuit with processimpurities.

174

MAJOR TECHNICAL PARAMETERS

MHR-100 SE
Parameters
Reactor heat capacity (MW)Useful electric power of generator (MW)Net power generation efficiency (%)
Helium temperature at reactor inlet/outlet (°C)
Helium flow rate through the reactor (kg/s)
Helium pressure at reactor inlet (MPa)Expansion ratio in turbine
Generator/TC rotation speed (rpm)MHR-100 SMR
Parameters
Reactor heat capacity (MW)Helium temperature at reactor inlet/outlet (°C)
Helium flow rate through the reactor (kg/s)
Helium pressure at reactor inlet (MPa)Steam-gas mixture pressure at HX inlet (MPa)HX-TCF 1
HX 1 capacity (MW)Helium/steam-gas mixture flow rate (kg/s)

Helium flow rate through turbine (kg/s)
Helium temperature at PCU inlet/outlet (°C)
SG power (MW)Helium flow rate through SG (kg/s)

Values
215
87.1
45.7
553 / 850
138
4.41
2.09
3000/
9000
126
850 / 558
22.3
12.1

Helium temperature at SG inlet/outlet (°C)
Steam capacity (kg/c)
Steam pressure at SG outlet (MPa)850 / 494
6.46
4.82

Steam-gas mixture temp. at inlet/outlet (°C)
HX-TCF 3
HX 3 capacity (MW)Helium/steam-gas mixture flow rate (kg/s)
Steam-gas mixture temp. at inlet/outlet (°C)

Steam-gas mixture temp. at inlet/outlet (°C)
HX-TCF 2
HX 2 capacity (MW)Helium/steam-gas mixture flow rate (kg/s)

Values215
450 / 950
81.7
5.0
5.3
31.8
12.1 /
43.5
350 / 650
58.5
22.2 /
60.9
350/750125
47.4/101
350/870

MHR-100 SMR

9. Design and Licensing Status
Optimization of reactor core design. Feasibility study of MHR-100-SMR plant application for large-scalehydrogen production, technical and economical evaluation of the plant potential to supply hydrogen to the
expected market. Studies of safety issues, with the emphasis on mutual influence of nuclear and hydrogenproduction components of the facility.

175

MAJOR TECHNICAL PARAMETERS
Parameters
Reactor heat capacity (MW)Helium temperature at reactor inlet/outlet (°C)
Helium flow rate through the reactor (kg/s)
Helium pressure at reactor inlet (MPa)IHX capacity (MW)Primary/secondary helium flow rate through IHX (kg/s)
Primary helium temp. at IHX inlet/outlet (°C)
Secondary helium temp. at IHX inlet/outlet (°C)
Secondary helium pressure at IHX inlet (MPa)Values215
300 / 750
91.5
5.0
217
91.5 / 113
750 / 294
230 / 600
5.50

MHR-100 OR

10. Fuel Cycle Approach
The MHR-100 fuel cycle approach is a once through mode without reprocessing. Fuel handling operations are
performed using the protective containers to avoid fuel assembly damage and radioactive product release.
Appropriately shielded containers are provided to protect the personnel against radiation impacts during
dismantling of the reactor unit components at fuel reloading. These measures are also applied at spent fuelmanagement.

11. Waste Management and Disposal Plan
Facilities for long-term storage of spent nuclear fuel (SNF) and solid/solidified radioactive waste (RW) are
included in the complex of a MHR-T commercial 4-unit NPP. The capacity of the designed SNF storage isdetermined from the condition of capability to store fuel unloaded from the NPP for 10 years. The capacity ofsolid/solidified RW storage facility is designed to provide storage of waste generated during the 10-year periodof NPP operation. After 10 years of storage at the NPP site, SNF and RW are to be removed for final
underground disposal. Radiochemical SNF reprocessing is considered as an option for future only.

12. Development Milestones
2014
2018
2020

176

Conceptual design completed
Feasibility study of plant application for large-scale hydrogen production
MHR-100-SMR is taken as the basis for near-term development of non-electricity
nuclear applications in Russia

AHTR-100 (Eskom Holdings SOC Ltd.,South Africa)
All content included in this section has been provided by, and is reproduced with permission of, Eskom SOC Ltd., South Africa.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type
Coolant/moderator

Eskom Holdings SOC Ltd.,South Africa
Modular high temperature gas
cooled reactor
Helium/graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

100 / 50

System pressure (MPa)9Core inlet/exit temperatures (oC)

406 / 1200

Fuel type/assembly array

Pebble bed with coatedparticle fuel~110 250 in core

Number of fuel assemblies

Forced circulation

Fuel enrichment (%)

LEU or WPu

Fuel burnup (GWd/ton)

86

Fuel cycle (months)

N/A; online / on-power
refuelling
Control rod insertion,negative temperature
coefficient
Passive

Main reactivity control
mechanismApproach to engineered safetysystems
Design life (years)

40

Plant footprint (m2)

To be confirmed

RPV height/diameter (m)11.4 / 6.05 (outer)
2.6 (inner)
0.4g PGA for main powersystem design
Initially once through
Uranium
Inherent safetycharacteristics; no core melt;high efficiency; small numberof safety systems
Concept design completed;
R&D activities in progress

Seismic design
Fuel cycle requirements /ApproachDistinguishing features

Design status

1. Introduction
When the PBMR® was defined in the mid-1990s, it was based on the industrially demonstrated technology of
the German Arbeitsgemeinschaft Versuchsreaktor (AVR) proof of concept reactor, and the German Thorium
High Temperature Reactor (THTR300) commercial scale reactor, integrated with a direct Brayton cycle helium
turbine power conversion unit, based on existing industrial gas turbines.
The PBMR® approach was to avoid any fundamentally new technologies and to move directly to the‘demonstration’ reactor, which was planned to also be a first of class of the commercial machine. While many
tests were done to confirm the performance that was achieved earlier, there were few new design elements,
except for the integration of the reactor with a helium gas turbine.
Given these potential advances available with current international technology, but not planned to be appliedby other programs, significant performance improvements could be achieved over the performance envisaged
for the original PBMR®.
In particular the use of carbon-composite materials to achieve higher operating temperatures (more than
1000C), and the use of molten salts as heat storage, would result in higher thermal efficiencies and better
flexibility. While all these technologies have been demonstrated at different scales in other industries, their177detailed application to High Temperature Reactor (HTR) designs would require an industrial-scale reactor
plant to prove their suitability for a commercial reactor. In order to implement these advances in ademonstration plant, the AHTR-100 was conceptualised by Eskom Holdings SOC Ltd. – PBMR SOC Ltd. iswholly owned by Eskom Holdings SOC Ltd.
With an output of temperature of 1200C the AHTR-100 is classified as a Very High Temperature Reactor(VHTR). Specific demonstrated nuclear technologies, such as the fuel design, will however remain the sameas that of the PBMR®.

2. Target Application
The AHTR-100 can produce electricity at high efficiency via a combined direct helium Brayton cycle andRankine bottoming cycle with an intermediate heat storage for load following or process heat applications ora bottoming steam cycle, as depicted in below Figure.

AHTR-100 schematic with topping, bottoming cycle and energy storage unit

3. Design Philosophy
As in the PBMR®, the AHTR-100 is a high-temperature helium-cooled, graphite moderated pebble bed reactor
but with a once-through fuelling scheme. The design safety targets and features means that the reactor can be
deployed close to the end user since there shall be no design base or credible beyond design base event thatwould need anyone living near the site boundary to take shelter or be evacuated. To achieve this goal there
shall be no need for engineered or moving mechanical components to ensure this target is met while theexposure to plant personnel shall also be significantly lower than today’s best international practice.

4. Main Design Features
Power Conversion System
A Brayton power conversion with direct gas turbine is adopted as topping cycle. It is a closed cycle where thehelium coolant is used to transport heat directly from the core to power turbine. The design incorporates asingle shaft for the turbine, the compressors and the power generator. Heat exchangers (up to 3) to remove
heat to the bottoming cycle is included.
From the reactor unit the hot helium enters directly to the turbine where energy is used to drive the shaft andtherefore the electric generator and compressors. From the turbine the helium then passes consecutively
through the primary side of the first high temperature heat exchanger, then the pre-cooler, the low pressurecompressor, intercooler, high pressure compressor and then on to the high-pressure intercooler before reentering the reactor unit.
The direct gas cycle is attractive since it promises the benefits of simplification, with the potential of lowering
the capital and operational costs. Due to the high outlet gas temperature, one will also expect a substantial
increase in the overall system efficiency.
178

This primary cycle will operate in baseload maximum capacity continually and provide 30% of the total plantelectricity. This limits reactor operating transients to startup, full load operation, and shut down.
The heat exchangers contain molten salt coolant in the secondary side removing heat from the primary circuitand storing it for use in a load following Rankine cycle.

Reactor Core and Fuel Characteristics
The core neutronic design results in a small cylindrical core with a diameter of 2.6 m. The effective cylindricalcore height is 9.35 m. In steady state (equilibrium core) operation the fuel sphere powers (average 0.91 kW)
and operational temperatures (1200oC) fulfil the design criteria. The core contains ~110 250 fuel spheres or‘pebbles’ with a packing fraction of 0.61. The fuelling scheme employed is the continuous on-line oncethrough method. Fresh fuel elements are added to the top of the reactor while used fuel pebbles are removed
at the bottom to keep the reactor at full power.
The fuel kernel (UO2) is coated by a first porous layer of pyrocarbon, followed by a dense layer of pyrocarbon,
a silicon carbide layer and an outer dense layer of pyrocarbon. About 13 330 coated particles and graphite
matrix material are made into an inner fuel zone and surrounded by a 5 mm outer fuel free zone to make up
the 6 cm diameter fuel sphere or pebble.

Reactivity Control
Excess reactivity is limited by once through, continuous refuelling cycle while adequate passive heat removal
ensures an inherent safe design with no event with significant fission product release being possible. Adequate
reactivity control and long-term cold shutdown capability are provided by two separate and diverse control rod
and small absorber sphere (SAS) systems while the overall negative reactivity temperature coefficient isnegative over the total operational range.

Fuel Handling System
Fuel spheres are circulated in the online handling system by means of a combination of gravitational flow andpneumatic conveying processes using helium at system operating pressure, as the transporting gas. The system
functions as an online fuel replenishing system. This involves fresh fuel replenishment, fuel unloading, anddischarging used fuel to the used fuel vessels.

Reactor Pressure Vessel and Internals
The average core height is 9.35 m and the reflector thickness 0.95 m. The side reflectors are manufactured
from nuclear grade graphite blocks that are stacked in columns to make up the geometry of the core. The side
reflector columns have borings for the control rods, as well as riser channels for the incoming coolant gas. All
the blocks are connected with graphite keys to prevent diversion of the coolant flow. The whole of theseceramic core internals is housed in a stainless-steel Core Barrel that is supported on the bottom of theprestressed concrete reactor pressure vessel.

5. Safety Features
The safety philosophy for modular HTRs has been described a number of times in the past 30 years and has
been adopted with a few modifications by AHTR-100 in the same manner as with the PBMR®. Its basis is thatan accident equivalent to severe core damage must be inherently impossible by limiting reactivity increases
and ensuring that decay heat can be removed passively after a loss of coolant event. The AHTR, like the
PBMR® has a simple design basis, with passive safety features that require no human intervention and thatcannot be bypassed or rendered ineffective in any way. If a fault occurs during reactor operations, the system,
at worst, will come to a standstill and merely dissipate heat through a heat pipe system on a decreasing curve
without any core failure or release of radioactivity to the environment.

(a) Engineered Safety System Approach and Configuration
The AHTR builds on the PBMR® nuclear reactor system that is designed to derive maximum safety benefits
from its inherent passive safety characteristics which are; designed to rule out core melt, all ceramics fuel,coated particle provides excellent containment for the fission product activity, large negative temperature
feedback, the helium coolant is chemically inert (single phase), large thermal capacity lead to slow thermaltransients, no common mode failure in the core (a single fuel failure does not lead to additional failures),ingress of water into core eliminated by design and air ingress limited.

(b) Decay Heat Removal/ reactor Cooling Philosophy
The Reactor Cavity Cooling System (RCCS) is a means to remove residual heat passively for a defined time,and indefinitely with the use of a passive heat pipe system. The use of a pre-stressed concrete pressure vesselin effect insulates the core from the atmosphere and as a result, the system requires passive heat removal by
the heat pipe system. In the event of the loss of active core cooling by the main circulation system, the heatpipe system is activated automatically through the temperature rise and are able to limit the increase in fueltemperature in the most affected region of the core to below the allowable fuel temperature limit.

179

(c) Containment Function
As with the PBMR®, the most important barriers to fission product release are the coatings of the fuel particles.
A second barrier is provided by the Helium Pressure Boundary. A third barrier is the confinement building.
The vented confinement is designed for very low leakage at low pressure, and to prevent damage tocomponents important to safety, as well as to contain the build-up of higher activity gas in the delayed phaseof a depressurisation event. Depending on the size of a pressure boundary break the system may be vented andthen closed again with the released gas filtered as required.

6. Plant Safety and Operational Performances
As in the PBMR®, the AHTR-100 safety does not rely on engineered systems that may fail but on the inherent
design and the laws of physics. The risk metrics core damage frequency and large early release frequency are
not applicable, but the same concepts are reflected in the immediate and delayed release category definitions.
The design of the AHTR based on the PBMR® represents a significant advancement in plant safety with anestimated delayed release category frequency of 1.0E-5 per reactor year while maintaining an expected
capacity factor of 95%.
The AHTR concept is directed to be a simplistic design, exhibit inherently safety characteristics and highoperational efficiency. The operating modes, states and transitions are under definition, but it is specified thatthe unit is able to shutdown with no human intervention requirements, in the event of LOFC.

7. Instrumentation and Control Systems
As in the PBMR®, the AHTR system consists of an inherently stable and slow acting heat source (Reactor
Unit), due to its large thermal capacity, which makes it nearly self-regulating, coupled to a fast-acting power
conversion machine. The Power Conversion Unit therefore require active control to remain stable under all
anticipated operating scenarios. The reactor power is adjusted by changes in the helium mass flow rate in the
power conversion unit. The helium inventory system is used to change the pressure (mass adjusted throughchanges in density) and power control is subsequently performed in combination with a bypass valve.

8. Design and Licensing Status
The design basis for the proof of concept machine has been completed for a direct cycle machine. The intent
is to test and proof several aspects of the technology prior to implementation in the commercial power plant.
The layout for the overall plant is being developed with operating modes, states and transitions progressively
defined.
The licensing framework for the proof of concept is also complete and the nuclear regulator is appraised on
the effort of the developments in the project. Reactor Plant Conceptual Phase has been completed with key
R&D work continuing in the field of qualifying materials and deisgn and construction of demonstration
components.

9. Fuel Cycle Approach
Once through uranium cycle is planned and analysed; pebble bed reactors are flexible to accommodate otherfuel cycles (plutonium or thorium) too.

10. Waste Management and Disposal Plan
The Waste Handling System would be based on the PBMR®-400 experience and designed to handle, store anddischarge low- and medium-level liquid and solid radioactive waste generated during normal operation,
maintenance activities, and upset events; including preparation for the final disposal. During final
decommissioning, the spent fuel spheres are removed from the interim storage and conveyed to a point where
they can be loaded into the Spent Fuel Transport Casks suitable for final disposal at a designated site.

11. Development Milestones
2010
2016
2017
2018
2019

180

PBMR® Project in care and maintenance since 2010.
AHTR-100 R&D activities commence.
AHTR-100 Version 1 concept completed.
R&D activities continue.
R&D put on hold pending funding availablility.

PBMR®-400 (PBMR SOC Ltd, South Africa)
All content included in this section has been provided by, and is reproduced with permission of, Pebble Bed Modular Reactor SOC Ltd., South Africa.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

Pebble Bed Modular Reactor SOC Ltd(PBMR®), South Africa
Modular high temperature gas cooled
reactorHelium/graphite
400 / 165
Forced circulation
9 (direct cycle / no secondary steam)
500 / 900
Pebble bed with coated particle fuelNumber of fuel assemblies~452 000 in core

Fuel enrichment (%)

9.6% LEU or WPu

Reactivity control mechanism
Approach to safety systems

Control rod insertion, negative
temperature coefficient
ActiveDesign life (years)

40

Plant footprint (m2)4200 (main structures only)

RPV height/diameter (m)30 / 6.2 (inner)

RPV weight (metric ton)

1250 (with vessel head)

Seismic Design (SSE)

0.4g PGA for main power system design

Fuel cycle requirements /ApproachDistinguishing features

Uranium once through; spent fuel stored
in tanks in the facility.
Inherent safety characteristics; no core
melt; high efficiency; small number ofsafety systems
Preliminary design completed; test
facilities demonstration; project stopped
in 2010. in care and maintenance.

Design status

1. Introduction
The Pebble Bed Modular Reactor (PBMR®) is based on the evolutionary design of the German HTR-Module
design. The PBMR® is designed in a modular fashion to allow for additional modules to be added in accordancewith demand. In addition, the PBMR® can be used as base-load station or load-following station and can be
configured to the size required by the community it serves.
Various reactor concepts have been under development since 1996. Most of these designs that are based on a
direct Brayton cycle aims for higher efficiencies. The maximum achievable power levels for the reactor wasincreased in several design steps to achieve a set target for installed cost/kW that would be comparable to coal
fired power when lifetime costs were evaluated. As a result, the design of the reactor core evolved from theoriginal base of 200 MW(t) adopted from the HTR-Module design to reach 400 MW(t) with an annular core.
Due to the world financial crisis in 2008 and short-term funding constraints, a reprioritization led to a decision
to concentrate on the electricity and process heat market with a single reactor product and thus a decision wasmade to use an indirect steam cycle. The direct cycle design was archived with a view to further progress thisdesign when conditions (financial and technology development in materials for the direct cycle) improve.

2. Target Application
The PBMR®-400 can produce electricity at high efficiency via a direct Brayton cycle employing a helium gas
turbine.

181

3. Design Philosophy
The PBMR®-400 is a high-temperature helium-cooled, graphite moderated pebble bed reactor with a multipass fuelling scheme. The design objectives and features mean that the reactor can be deployed close to theend user since there shall be no design base or credible beyond design base event that would need anyone
living near the site boundary to take shelter or be evacuated. To achieve this objective there shall be no needfor engineered or moving mechanical components to ensure this objective is met while the exposure to plant
personnel will be significantly lower than today’s best international practice.

4. Main Design Features
Power Conversion System
The Brayton cycle power conversion cycle with direct gas turbine is adopted. It is a closed cycle where thehelium coolant is used to transport heat directly from the core to the power turbine. The design incorporates
turbine, compressors and power generator in a single shaft. The flow of the Helium is depicted in the figure
below. The direct gas cycle is attractive since it promises the benefits of simplification, with the potential oflowering the capital and operational costs. Due to the high outlet gas temperatures, one will also expect a
substantial increase in the overall system efficiency.

Reactor Core and Fuel Characteristics
The core neutronic design is an annular core with an outer diameter of 3.7 m and an inner diameter of 2 m
shaped by the fixed central reflector. The effective cylindrical core height is 11 m. In steady state (equilibriumcore) operation the fuel sphere power (maximum 2.7 kW per sphere) and operational temperatures (<11000C)
fulfil the design criteria set. The core contains ~452 000 fuel spheres or ‘pebbles’ with a packing fraction of
0.61. The fuelling scheme employed is a continuous on-line multi-pass system. Fresh fuel elements are added
to the top of the reactor while used fuel pebbles are removed at the bottom to keep the reactor at full power.
On average fuel spheres are circulated six times through the reactor. This reduces power peaking and maximum
fuel temperatures in normal operation and loss of coolant conditions.
The coated particle pebble fuel used is shown below. The fuel kernel (UO2) is coated first by a porous layer ofpyrocarbon, followed by a dense layer of pyrocarbon, a silicon carbide layer and an outer dense layer ofpyrocarbon. About 15 000 of these coated particles and graphite matrix material are made into an inner fuelzone and surrounded by a 5 mm outer fuel free zone to make up the 6 cm diameter fuel sphere or pebble.

Reactivity Control
Excess reactivity is limited by continuous refuelling while adequate passive heat removal ensures an inherent
safe design with no event with significant fission product release being possible. Adequate reactivity control
and long-term cold shutdown capability is provided by two separate and diverse systems while the overall
reactivity temperature coefficient is negative over the total operational range. The reactivity control systemfacilitates load following between 40% and 100%.

Fuel Handling System
Fuel spheres are circulated in the online handling system by means of a combination of gravitational flow andpneumatic conveying processes using helium at system operating pressure, as the transporting gas. The system
functions as an online fuel replenishing system. This involves fuel unloading, discharging spent anddamaged/worn fuel to used fuel vessels, reloading fresh fuel and fuel that can be returned back to the reactor.

Reactor Pressure Vessel and Internals
The average core height is 11 m and the annulus thickness is fixed at 0.85 m. The centre reflector diameter is
2 m and contains eight borings for the Reserve Shutdown System, consisting of borated graphite spheres of 10
mm diameter. The centre and side reflectors are manufactured from nuclear grade graphite blocks that arestacked in columns to make up the geometry of the core. The side reflector columns have borings for thecontrol rods, as well as riser channels for the incoming coolant gas. All the blocks are connected with graphite
keys to prevent diversion of the coolant flow. The whole of these ceramic core internals is housed in a stainless182

steel Core Barrel that is supported on the bottom of the Reactor Pressure Vessel.

PBMR®-400 power conversion unit components and layout, and He flow

5. Safety Features
The safety philosophy for modular HTRs has been described a number of times in the past 30 years and has
been adopted with a few modification by PBMR®. Its basis is that an accident equivalent to severe core damage
must be inherently impossible by limiting reactivity increases and ensuring that decay heat can be removed
passively after a loss of coolant event. The PBMR® has a simple design basis, with passive safety features that
require no human intervention and that cannot be bypassed or rendered ineffective in any way. If a fault occurs
during reactor operations, the system, at worst, will come to a standstill and merely dissipate heat on adecreasing curve without any core failure or release of significant radioactivity to the environment.

Engineered Safety System Approach and Configuration
The PBMR® nuclear reactor system is designed to derive maximum safety benefits from its inherent passive
safety characteristics which are; designed to rule out core melt, all ceramics fuel, coated particle provides
excellent containment for the fission product activity, large negative temperature feedback, Helium coolant ischemically inert (single phase), large thermal capacity lead to slow thermal transients, no common mode failurein the core (a single fuel failure does not lead to additional failures), ingress of water into core eliminated bydesign and air ingress limited.

Decay Heat Removal/ reactor Cooling Philosophy
The Reactor Cavity Cooling System provides a means to remove residual heat passively for a defined time,and indefinitely with the use of an active system after refilling the cooling system. For this to work, the Reactor
Pressure Vessel and the core need to be long and slender. The belt region of the RPV is not insulated to allow
heat radiation and convection to the water filled cavity cooler. In the event of the loss of active core coolingby the main circulation system, the cavity cooler and/or the building structural materials are able to limit theincrease in fuel temperature in the most affected region of the core to below the allowable fuel temperature
limit.

Containment Function
The most important barriers to fission product release are the coatings of the fuel particles. A second barrier isprovided by the Helium Pressure Boundary. A third barrier is the confinement building. The vented
confinement is designed for very low leakage at low pressure, and to prevent damage to components importantto safety, as well as to contain the build-up of higher activity gas in the delayed phase of a depressurisation
event. Depending on the size of a pressure boundary break the system may be vented and then closed again
with the released gas filtered as required.

6. Plant Safety and Operational Performances
The PBMR®-400 safety does not rely on engineered systems that may fail but on the inherent design and thelaws of physics. The risk metrics core damage frequency and large early release frequency are not applicable,
but the same concepts are reflected in the immediate and delayed release category definitions. The design of183

the PBMR® represents a significant advancement in plant safety with an estimated delayed release category
frequency of 1.0 x 10-5 per reactor year while maintaining an expected capacity factor of 95%.

7. Instrumentation and Control Systems
The PBMR® system consists of an inherently stable and slow acting heat source (Reactor Unit), due to its large
thermal capacity, which makes it nearly self-regulating, coupled to a fast-acting power conversion machine.
The Power Conversion Unit therefore require active control to remain stable under all anticipated operating
scenarios. The reactor power is adjusted by changes in the helium mass flow rate in the power conversion unit.
The helium inventory system is used to change the pressure (mass adjusted through changes in density) andpower control is subsequently performed in combination with a bypass valves.

8. Plant Arrangement
Nuclear Island
Conventional Island

Citadel

PBMR®-400 building layout

9. Design and Licensing Status
The Reactor Plant Preliminary Design was completed and demonstration of key technologies were underway
when the project was terminated in 2010.

10. Fuel Cycle Approach
Once through uranium cycle was planned and analysed; pebble bed reactors are flexible to accommodate otherfuel cycles (plutonium or thorium) too.

11. Waste Management and Disposal Plan
The Waste Handling System is designed to handle, store and discharge low- and medium-level liquid and solidradioactive waste generated during normal operation, maintenance activities, and upset conditions of the
PBMR®-400; including preparation for the final disposal. During final decommissioning, the spent fuel spheres
are removed from the Spent Fuel tanks and conveyed to a point where they can be loaded into the Spent Fuel
Transport Casks suitable for final disposal at a designated site.

12. Development Milestones
1993
1995
1999
2002
2002
2004
2006
2006
2007
2009
2010
2018

184

The South African utility Eskom identifies PBMR as an option for new generating
capacity.
Start of the first pre-feasibility study.
Design optimization: PBMR®-268 with dynamic central column.
Design changed to PBMR®-400 with fixed central column.
The Pebble Bed Micro Model (PBMM) demonstrated the operation of a closed, three
shaft, pre- and inter-cooled Brayton cycle with a recuperator.
Vertical layout of turbo machines changed to conventional single horizontal layout.
Commissioning of Helium Test Facility for full scale system and component tests.
Tests starts in the Heat Transfer Test Facility.
Advanced fuel coater facility commissioned.
Coated particles sent for irradiation testing at INL.; Alternative process heat markets anddesigns explored.
Project closure.
Project in care and maintenance.

HTMR100 (STL Nuclear (Pty) Ltd, South
Africa)All content included in this section has been provided by, and is reproduced with permission of, STL Nuclear (Pty) Ltd, South Africa.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of
STL Nuclear (Pty) Ltd , SouthoriginAfricaReactor type
(HTGR) Pebble Bed
Coolant/moderator
Helium, graphite
Thermal/electrical capacity,
100 / 35 single module plant
MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation-4 / 16Click or tap here to enter text.

Core inlet/outlet coolant
temperature (oC)

250 / 750

Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)
Refuelling cycle (months)
Core Discharge Burnup (GWd/ton)Reactivity control mechanism
Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)

TRISO particles in pebbles; LEU/Th~150 000; around 125 to 150
pebbles/day throughput
10
Online fuel loading
80-90
Control Rods in the reflector
Passive40
5 000 (buildings only)
15.7 / 5.6 (flange outer diameter)
155
0.3 g (generic site), 0.5 g underconsideration)
Various options (see below)
No core meltdown, no activeengineered safety systems
Basic design

Fuel cycle requirements/approachDistinguishing features
Design status

1. Introduction
The HTMR100 (High Temperature Modular Reactor) pebble bed reactor is a high temperature gas cooled
reactor, graphite moderated and cooled by forced helium flow. The existing design of the module is to
produce high quality steam which is coupled to a steam-turbine/generator system to produce 35 MW(e). The
steam can be used in a wide range of cogeneration applications. The reactor is also suitable to provide direct
high temperature energy for process heat. The design of the reactor is based on proven technology andtherefore no new basic technology development is needed. The size of the reactor and the fuel cycle werechosen to simplify the design and operation of the module. The approach to small intrinsic safe modular
units ensures continuous production, easy road transportability, skid mounted sub systems, wider range ofmanufactures, fast construction and an enhanced licensing process.

2. Target Application
The HTMR100 can supply electric power to any distribution grid and to standalone or isolated electricity
users. It can be deployed as single modules or multi-module plants as well as for medium temperatureprocess heat applications (later also upgradable to very high temperature). The HTMR100 is a perfect fit forclients who want to progressively extend their generating capability. The unique safety characteristics makeit possible to introduce and construct these plants to non-nuclear countries. First-world countries that want toutilize their stock of Plutonium for peaceful applications are also markets for HTMR100 reactors.

3. Design Philosophy
The reactor has good load following characteristics which is needed for stand-alone (not grid coupled)
applications. The ‘Once Through Then Out’ (OTTO) fuelling scheme leads to a simple and cost-effective
fuel management system. The relatively low primary loop pressure requires a thinner walled pressure vessel
185

and thus an easier manufacturing process, resulting in a wider range of vessel manufacturers. The HTMR100
plant design caters for different site and client requirements. It allows flexibility in protection against
external events and flexibility in multi module configuration and power capacity.

4. Main Design Features
Power Conversion
The flow through the core is from top to bottom where theheated gas is collected in a hot plenum. From the plenum the
hot gas flows through a connecting pipe to the steam
generator. The power conversion system uses a helical coilsteam-generator unit supplying super-heated steam to theturbine. The main system will be supplied as four skid
mounted units namely the condenser, turbine, gearbox andelectric generator. The turbine can be used in a back-pressure
configuration or intermediate temperature steam can be taken
off for process heat applications.

Reactor Core

Reactor core and power conversion layout

The reactor unit consists of a steel pressure vessel, a steel core barrel, graphite reflector blocks, neutron
absorber rods, rod guide tubes, drive mechanisms and in-vessel instrumentation. The vessel is designed for 4
MPa pressure. The graphite structure allows for differential expansion and volumetric changes due totemperature and neutron fluence induced distortion. This is done to keep the stresses low and minimize
primary fluid bypass and leaks. The side, top and bottom reflector material is nuclear grade graphite.

Fuel Characteristics
The fuel elements (FE) for the HTMR100 are 60 mm diameter spheres consisting of a spherical fuel zone ofapproximately 50 mm diameter, in which the TRISO-coated particles are randomly distributed in thegraphitic matrix material. A fuel-free shell of graphite matrix of about 5 mm in thickness is then moulded to
the fuel zone. The fuel kernel and coatings serve as a fission product barrier in normal and accident operating
conditions. There are various types of fuel that will be used in the HTMR100 reactor (see below for details).
A Fuel Qualification and Test programme will be conducted on the fuel prior to loading of the reactor. The
HTMR100 operates on a much longer burn-up fuel cycle compared to conventional reactors. The nonproliferation characteristic of the OTTO cycle is the extended time the pebbles reside inside the core, making
it more difficult to divert partially burnt fuel.

Fuel Handling System
A six-month supply of fresh fuel is kept in the fresh fuel storage facility. New spherical fuel elements (freshfuel) are loaded by the fuel loading machine into a charge lock. The charge lock is purged, filled with clean
helium and pressurised to system pressure, before it is opened, and fuel is gravity fed into the core cavity.
The charge lock has a physical capacity for approximately one full-power day’s fuel sphere inventory.

Reactivity Control
Eighteen neutron absorber rods are provided in graphite sleeves inside the graphite side reflector blocks. The
absorber rods can be operated independently as a group or as sub-groups, as required by the reactor operating
control system. A control rod consists of several rod absorber material segments, pinned together to form
articulating joints. The segments consist of sintered B4C absorber material, sandwiched between an innerand an outer tube segment. The inner tube segment allows cooling helium gas to flow form the top down inthe circular channels. Each rod is equipped with a position indicator which measures the position of the rod
over its entire positioning range and with position indicators for the upper and lower limit positions.

Reactor Pressure Vessel and Internals
The Reactor Pressure Vessel (RPV) is constructed in compliance with the ASME III subsection NB code. It
comprises two main components reactor of vessel body and vessel head which is bolted to vessel body. The
reactor vessel body consists of several forged ring-components circumferentially welded together.
The core structures consist of the metallic parts and the graphite structures. The function of these internal
structures is to provide stable core geometry, neutron reflection, cold and hot gas channelling, fuel element
flow, shielding, thermal insulation and support of the control and shutdown systems guide tubes and theneutron source. The functional design of the structural core internals is such that they can withstand thesteady state and transient loadings during normal operation, anticipated operational occurrences and designbasis accidents.
The shape and structure of the inner side reflector wall and the 30° angled core bottom permit uniform fuelelement flow. The loads borne by the ceramic internals are transferred to the steel core barrel and then to thereactor pressure vessel through metallic components such as the lower support structure and the core barrel
axial and radial supports.
186

All areas of the core internals are designed for the service life of the reactor. Access for ceramic structure
inspections can be done through the fuel loading channel and the reflector rod holes.

Reactor Coolant System
Helium enters the outer part of the co-axial duct and flows up the helium risers, it is then directed into thepebble bed core and removes heat as it flows downwards. The helium then collects in the lower plenum andis directed in the inner pipe of the co-axial duct back to the steam generator. Electric blowers are used tomove the helium within the reactor pressure boundary. The secondary power conversion loop uses a
conventional Rankine cycle and heat is transferred from the helium loop to the water loop to form steamwhich in turn drives the turbine.

5. Safety Features
Engineered Safety System Approach and Configuration
In principle the plant is designed to perform its safety functions without reliance on the automated plant
control system, or the operator. The engineered safety system of the plant has no engineered safety systems
in terms of active human or machine intervention to assure nuclear safety. Provision for beyond design basis
conditions is made. Beyond design basis scenarios include the non-functioning/non-insertion of all active
control and shutdown systems. The reactor core characteristics e.g. small excess reactivity and strong
negative reactivity coefficient with temperature will shut down the reactor and maintain a condition where nodamage to the fuel, core structures and reactor vessel occurs. Excessive reactivity increases during water orwater vapor ingress (increasing moderation) is prevented by designing the reactor for limited heavy metal
content of the fuel.

Decay Heat Removal System / Reactor Cooling Philosophy
The Reactor Cavity Cooling System (RCCS) removes heat radiated from the reactor towards the reactor
cavity walls. It consists of welded membrane tubes arranged side-by-side on the inside of in the reactor
cavity wall. Water is circulated through the tubes to form a cold wall. The RCCS is a passive system andconsists of three independent cooling trains and is designed for all postulated design basis conditions.

Spent Fuel Cooling Safety Approach / System
The spent fuel will be stored in special tanks of which the height/diameter is in excess of 4. Thermosiphons(heat pipes) are attached to the outside walls of these tanks which removes heat to the outside of the spent
fuel area. The condenser ends of these heat pipes are then fitted with fins to dissipate the heat to atmospherein an entirely passive way. There is also the possibility to use Stirling engine/generators on the condenser
ends to generate electricity for charging batteries and in so doing provide a measure of energy conservation.

Containment System
The primary fission product barrier is the SiC layer of the TRISO coated fuel particles, which keeps thefission products contained under all postulated events, even if the second barrier (the primary pressure vessel
system) and the third barrier (the building filter system) fails.

Chemical Control
The probability of an in-leakage of water in case of a tube rupture in the steam generator is reduced by its
vertical positioning relative to the hot parts of the reactor core. Water leaking from a ruptured tube will thenaccumulate in the shell of the steam generator and consequently will not be able to enter the core inquantities that will cause damage. In case of air ingress due to a break in the pressure boundary will be aslow diffusion process against the outflow rate of helium coolant. In addition, the construction of the coolant
flow paths in and around the core structures is such that a naturally driven chimney type flow of oxygen is
avoided inherently. Also, the grade of graphite used in the fuel and core structures is of such purity that selfpropagating oxidation reactions are naturally not possible.

6. Plant Safety and Operational Performances
The central consideration is the demand for high availability of process steam supply and/or electricity
generation. To reduce or minimize the NSSS daily or weekly load changes of the reactor, the preference is tochange the ratio between steam supply and electricity supply. Excess steam and/or electricity can be utilized
in the desalination plants to provide water as a sellable commodity earning additional revenue. This allows
the plant to operate virtually continually at full power very close to the plant availability.

7. Instrumentation and Control System
The Automation System (ATS) comprises that group of safety and non-safety C&I systems that provide
automated protection, control, monitoring and human-system interfaces. The three specific systems in the
HTMR100 system structure define control and instrumentation are plant control, data and instrumentation
system, equipment/investment protection system and protection system.

187

8. Plant Layout Arrangement
The reactor building contains the safety equipmentthat provides the necessary functions for the safe
shutdown of the reactor under all design basis
conditions. The reactor building is partiallysubmerged below ground level such that thereactor and steam generator cavities are completely
protected against postulated external threats. The
depth can be further adapted to suit the geologicalconditions of the specific site to provide for thenecessary level of seismic protection.

Plant Layout

The reactor building, electrical building and auxiliary buildings are connected by means of undergroundtunnels, providing protection for interlinked services and, it also ensures that spent fuel is never brought
above ground level. Provision is made for the storage of all spent fuel produced during the operating life of
the plant. The reactor building is seismically designed to withstand a design basis earthquake (DBE) andtogether with the spent fuel storage bunker, is the only safety related building structure of the HTMR100.
The Turbine Building provides the foundation and housing for the Power Conversion System, including
other support systems such as the compressed air, water sampling, HVAC, Voltage Distribution Systems,
permanent 11kVAC and 400VAC diesel generator sets and steam safety valves.
The Electric Building houses the main control and computer rooms, primary and secondary plant security
alarms rooms and provides the primary access facilities for the nuclear island and the energy conversion area.
This centre also provides space for activities associated with plant administration and security services. The
plant control, data, and instrumentation system control/display panels and computers are housed in thecontrol room.

9. Testing Conducted for Design Verification and Validation
Under design.

10. Design and Licensing Status
Conceptual design is in an advanced stage. The core neutronic, thermo-hydraulic and heat transfer analyses
are being done to optimize the performance and verify the safety analysis. Nuclear Regulator engagement
was initiated with the aim of commencing the pre-assessment for licensing in order to reach design
certification status at the end of the conceptual phase.

11. Fuel Cycle Approach
The reactor design can accommodate various fuel types with different fuel cycles, including mixtures ofthorium and plutonium or thorium and uranium. Studies have shown the reactor can utilise: (i) 10% LEU (710g HM/sphere); (ii) Th/LEU mix of 50% LEU (20% enriched) and 50% Th (10-12g HM/sphere); (iii)
Th/HEU mix with 10% HEU (93% enriched) and 90% Th (10-12g HM/sphere); and (iv) Th/Pu mix with
15% reactor grade Pu by mass, (12g HM/sphere). Reprocessing of the HTMR100 fuel elements is notintended.

12. Waste Management and Disposal Plan
The HTMR100 fuel elements can be stored and disposed of a fuel sphere but available technology needs tobe assessed for volume reduction. Disposal of spent spherical fuel elements from the HTMR100 is executedin the following sequence: (i) Direct transfer of spent fuel elements into a flask inside the cast iron highenergy spent fuel casks (Hi-cask); (ii) Immediately after filling the H-cask they are sealed and transferred tothe spent fuel cool-down facility on site; (iii) Once cooled down, the flask filled with fuel is transferred fromthe H—cask to a low energy spent fuel concrete cask (Low-Cask); iv) The Lo-Cask is transported to the low
energy on-site interim storage facility; v) For offsite transport the flask is transferred to a shipping/transport
cask for shipping to an ultimate repository. Approximately 55 000 fuel elements will be discharged for on
full-power year of operation and only one or two flasks containing physically damaged fuel spheres, singled
out by the fuel unloading machine (failed fuel separator), may be required in the lifetime of the core. As innormal and accident conditions the coated particles maintain its excellent fission product retention
capabilities and fission products are almost entirely retained within the fuel element kernels. Also, the releaseof these nuclides into the cask or flask atmosphere from the number of fuel element particles with defective
SiC coatings is very low.

13. Development Milestones
2012
2020
2021
2022
188

Project Started
Preparation for pre-licensing application
Concept design completed
Basic design started

Energy Multiplier Module (EM2) (General
Atomics, United States of America)
All content included in this section has been provided by, and is reproduced with permission of, General Atomics, USA.
Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

500 / 265

NSSS operating pressure(primary/secondary), MPa
Core inlet/outlet coolant
temperature (°C)
Fuel type/assembly array

13.3

Number of fuel assemblies in thecore
Fuel enrichment (%)

Elevation view of EM2 modular building element
employing two modules on a single seismicallyisolated platform

General Atomics (GA), UnitedStates of America
Modular high temperature gascooled fast reactor
Helium / none

Forced cooling

550 / 850
Uranium carbide pellet/
hexagonal
85
14.5

Refueling cycle (months)

360

Core discharge burnup (GWd/ton)130

Reactivity control mechanism

Control rod drive mechanism

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

90 000 (four modules)

RPV height/diameter (m)12.5 / 4.6

RPV weight (metric ton)

700

Seismic design (SSE)

0.3 g

Fuel cycle requirements/approach

Open fuel cycle

Distinguishing features

Convert-and-burn/Siliconcarbide composite cladding/Fission gas collection
Conceptual design

Design status

1. Introduction
Energy Multiplier Module (EM2) is a helium-cooled fast reactor with a core outlet temperature of 850°C. It is
designed as a modular, grid-capable power source with a net unit output of 265 MW(e). The reactor converts
fertile isotopes to fissile and burns them in situ over a 30-year core life. EM2 employs a direct closed-cycle gas
turbine power conversion unit (PCU) with a Rankine bottoming cycle for 53% net power conversion efficiency
assuming evaporative cooling. EM2 is multi-fuel capable, but the reference design uses low-enriched uranium(LEU) with depleted uranium (DU) carbide fuel material with silicon carbide (SiC) composite cladding
material (i.e., SiGA®).

2. Target Application
The EM2 is being developed for the electricity generation and high temperature use.

3. Design Philosophy
EM2 design philosophy is to develop a new nuclear plant to address the following goals for enhancing the
likelihood of commercial success: i) economic parity with fossil fuel generation in the US, ii) improved siting
flexibility via dry cooling and site accessibility, iii) passive safety for sustained protection during long-term
station blackout and other severe accidents, and iv) improved fuel resource utilization, reduced nuclear waste,and high proliferation-resistance.
1894. Main Design Features
The EM2 core was specifically designed to extend the fuel burnup to maximize the fuel utilization with areasonable amount of initial uranium loading. From this perspective, a fast neutron spectrum was chosen. For
the thermal efficiency of the plant, high temperature operation was chosen. These design choices require useof high temperature material for the fuel and core structure. To accommodate high fuel burnup, the fission
gases are removed from the fuel and stored in a collection system, which maintains the pressure in the fuelslightly lower than the primary system pressure.

Power Conversion
The power conversion is based on a combined cycle with a direct helium Brayton cycle and a Rankine
bottoming cycle. The helium Brayton cycle is provided by located in the PCU, while the Rankine cycle is
implemented in the facility outside the reactor building. The Brayton cycle incorporates the turbo-compressor(T/C) and generator, which are mounted on an in-line vertical shaft suspended by active magnetic bearings.
The cycle also incorporates two heat exchangers (HX), a recuperator and a precooler. The generator uses apermanent magnet (PM) rotor to eliminate losses associated with a wound rotor and exciter.

Reactor Core
The core is supported by the support floor through the core barrel, which is attached to the vessel below thecross-duct. The upper carbon-composite (C-C) heat shield protects the top head elements from the hot helium.
The vessel is internally insulated with silica/alumina fibrous insulation retained with a C-C cover plate. To
achieve high fuel utilization, the EM2 core utilizes the “convert and burn” concept, in which the core is divided
into fissile and fertile sections. The fissile section is the “critical” section at beginning-of-life. It contains~14.5% LEU to sustain the chain reaction and provide excess neutrons to convert DU fertile to fissile material.
The average enrichment of the total active core is 7.7%. The reflector consists of an inner section of
zirconium-alloy blocks and an outer section of graphite blocks. The basic building block of the EM2 fuelsystem is the hexagonal assembly, of which there are 85 in the core. Eighty-one assemblies are joined into 27
tri-bundles and four remain as individual assemblies.

Fuel Characteristics
Uranium carbide (UC) is used to meet the high uranium loading requirement; UC has a very high thermal
conductivity, is compatible with the SiGA cladding, and has a suitably high melting point. Each annular fuelpellet is a sintered “sphere-pac” with a specified interstitial and internal distributed porosity to allow for fastermigration of volatile fission products. SiC composite is especially attractive due to its stability under long-term
irradiation as demonstrated in a multi-year irradiation campaign. Both the fuel and cladding materials meet
design criteria temperature limits for both normal operations and accident conditions.

Fuel Handling System
The core is accessed by a refueling machine from the maintenance hall floor. An articulated arm extends
through the containment and reactor vessel penetration to select and withdraw a tri-bundle assembly and load
it into a sealed, air-cooled storage container. The container is moved to the end of the maintenance hall, where
it is lowered into the fuel storage facility. This facility has the capacity for 60 years of operation. The spent
fuel is cooled within the sealed containers by passive natural convection of air. No water or active cooling isrequired.

Reactivity Control
Reactivity control is provided by the 18 control rods and 12 shutdown rods. Both control rod system andshutdown rod system have sufficient negative reactivity to render the core cold subcritical. The control andshutdown drives are located at the top of the reactor vessel. The control rod drives use a ball-screw drive, while
the shutdown rods use linear motors.

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) is constructed from welded ring-forgings or rolled plate. The vessel contains
large penetrations for the two cross-vessels and a flanged hatch. The top head has the penetration for refuelingaccess and control elements. The RPV has no external insulation but is internally insulated. This insulation
maintains the vessel well below 371°C during normal operation and design basis accidents, which allows theuse of SA-533 grade B material.

Reactor Coolant System
The reactor coolant system (RCS) encloses the reactor and PCU and includes all subsystems that contain,transfer, regulate, purify, monitor, and store primary coolant helium. The RCS provides controlled pressurized
helium for transport of heat from the reactor to the PCU during normal operation and from the reactor to thedirect reactor auxiliary cooling system (DRACS) during shutdown conditions when the PCU is not available.
It also provides a Safety Class I barrier against the release of radioactivity to the containment during normal
and abnormal conditions.

190

5. Safety Features
The EM2 safety design uses a defense-in-depth approach, which employs three successive, encompassing
barriers against the release of radionuclides. Each barrier relies ultimately on passive means for protection of
its integrity for normal and abnormal operation. The first barrier is the SiGA cladding. The second barrier isthe primary vessel system, which encompasses the reactor, PCU, and DRACS. The third barrier is thefree-standing, below-grade containment. Because the fuel is vented, the fission gas collection system (FGCS)
is an extension of the first barrier.

Engineered Safety System Approach and Configuration
The engineered safety features are designed to detect incipient conditions that pose a threat to the fission
product barriers and initiate protective actions including reactor shutdown, sustained heat removal from thecore, and protected integrity of the primary pressure boundary and/or containment. EM2 incorporates several
systems classified as engineered safety features, emergency systems or systems with safety significance suchas DRACS, containment system, reactivity control system, reactor protection system, Class 1E Alternating
Current (AC) and Direct Current (DC) power systems, and fission product vent system (FPVS).

Decay Heat Removal System/Reactor Cooling Philosophy
The DRACS safely removes core decay heat during normal shutdown and accident conditions when the PCU
is not available. It provides controlled core heat removal during an anticipated transient without scram (ATWS)
and active core heat removal during special maintenance conditions characterized by need for low temperatures
and/or low helium pressures.

Spent Fuel Cooling Safety Approach/System
The spent fuel storage facility (SFSF) accommodates 60 years of spent fuel storage. The SFSF provides
adequate passive dry cooling of the spent fuel, protection from external threats, and monitoring for spent fuelstorage canister (SFSC) leakage. The SFSF is located below grade and is protected by an extension of theprotective shield that covers the reactor building maintenance hall. The SFSF roof is aligned with themaintenance floor to allow easy transport of the SFSCs to the SFSF. The SFSF has redundant, elevated air
intakes and outlets for cooling.

Containment System
The primary heat transport system (PHTS) is enclosed by a sealed, below-grade containment, which is divided
into three connected chambers with structural ligaments around the reactor chamber that also serve as shielding
to all access to the two side chambers. The containment is hermetically sealed with an inert (argon) atmosphereat ~20 psig (0.14 MPa). The peak pressure rating is 90 psig (0.62 MPa). The design leakage rate is less than
0.2% per day.

Chemical Control
Helium coolant is maintained at a high level of purity by the helium purification system (HPS). The HPS
maintains the total oxidant (O2, CO, H2O, CO2) concentration in the helium below 0.35 ppmv and N2 and H2concentrations below 10 ppmv; at these levels of purity, chemical interactions between coolant contaminants
and the fuel are not life limiting. The high temperature adsorber (HTA) is shared with the FPVS. The HTA
removes halides and metals. The effluent from the HTA is transported to the reactor auxiliary building, where
the remaining oxidants and noble gases are removed.

6. Plant Safety and Operational Performances
Unlike large power reactors, each EM2 module utilizes a unique, non-synchronous, variable speed generator
with a frequency inverter and generator load commutator to follow load demand. In the automatic load
following mode, the generator speed is set by modulating the generator speed. A field-oriented control
algorithm in the frequency converter controls the generator torque that decreases or increases the generator
speed, which in turn determines helium flow. The ability to control the turbo-generator speed through field
control replaces traditional mechanical control elements with digital electronic control. An advantage of thevariable speed control is that it maintains primary system structures at near constant temperature with load sothat rapid load following should be possible.

7. Instrumentation and Control Systems
EM2 deploys advanced sensors for high operating temperature condition such as solid-state and SiC neutron
flux monitors. The plant control including startup, operation, and shutdown is conducted through integrated
control system actions, which regulate reactor power and turbomachinery to respond to the plant transients.
Plant control functions are performed by the power control and the process component control systems. The
power control system includes control rod drive mechanism (CRDM) and reactor coolant flow. The processcomponent control system includes a non-synchronous, variable speed generator and a frequency inverter.

191

8. Plant Layout Arrangement
The baseline EM2 plant is composed of four 265 MW(e) modules for a combine net power of 1060 MW(e) toa utility grid for evaporative cooling and 960 MW(e) net for dry cooling. Each module consists of a complete
powertrain from reactor to heat rejection such that the modules can be built sequentially and operated
independently. The plant layout covers 9.3 hectares (23 acres), not including the switchyard. The maintenancehall floor is at grade level, and the roof serves as a protective shield structure. The maintenance hall serves all
four reactors.

EM2 plant layout on 9 hectares of land

9. Testing Conducted for Design Verification and Validation
The design verification and validation plan will be developed. The experimental measurements have been
partially conducted for i) UC spherical kernel irradiation tests in High Flux Isotope Reactor (HFIR) Cycle 490
in 2021 and measurements of fission gas release and swelling in 2022, ii) a series of thermal, mechanical, andsafety measurements of the SiC composite tube, and iii) a low-fluence irradiation of zirconium silicide (Zr3Si2)
reflector material in 2016.

10. Design and Licensing Status
Interaction with the Nuclear Regulatory Commission (NRC) was initiated for the fuel qualification plan. The
licensing process of the EM2 prototype plant was not launched. It will be licensed using the two-step 10 CFRPart 50, and follow-on commercial plants will be licensed under the one-step 10 CFR Part 52 process. The
technology-inclusive, risk-informed, performance-based licensing framework developed by the Licensing
Modernization Project will be used to develop the NRC license applications.

11. Fuel Cycle Approach
The EM2 open fuel cycle with LEU/DU vented fuels exceeds 30 years without refueling or shuffling, leading
to a reduced cost of power, low proliferation risk, high fuel utilization, and low mass of waste streams. The
core is capable of burning used light water reactor (LWR), plutonium, and thorium fuels. Fissile self-sufficientfuel cycle is feasible by removing fission products from the EM2 spent fuel.

12. Waste Management and Disposal Plan
The radioactive waste handling system collects, transfers, and stores radioactive materials from plant operating
systems, including immobilized fission gases from the vented fuels. The spent fuels are stored in a storage
facility at the site. After cooling, the used fuel will be directly disposed of or recycled.

13. Development Milestones
Phase I
Phase II

Phase III

192

Conceptual design of reactor core and power plant: high risk research and development
Demonstration unit development: demonstration of major components; fuel design andqualification; high temperature material development
Prototype unit development: demonstration of plant operation; licensing for commercial
unitFast Modular Reactor (FMR) (General
Atomics, United States of America)
All content included in this section has been provided by, and is reproduced with permission of, General Atomics, USA.
Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type
Coolant/moderator

Reactor building and major systems andcomponents arrangement designed with defense in
depth

General Atomics (GA),
United States of America
Gas-cooled fast reactor(GFR)
Helium / none

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

100 / 50NSSS operating pressure(primary/secondary), MPa
Core inlet/outlet coolant
temperature (°C)
Fuel type/assembly array

7

Forced cooling

509 / 800

Number of fuel assemblies in the core

Uranium dioxide pellet/
hexagonal
198

Fuel enrichment (%)

19.75

Refueling cycle (months)

96

Core discharge burnup (GWd/ton)100

Reactivity control mechanism

Control rod drive mechanism

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

38 000 (four modules)

RPV height/diameter (m)12 / 5

RPV weight (metric ton)

700

Seismic design (SSE)

0.3 g

Fuel cycle requirements/approach

Open fuel cycle

Distinguishing features

Silicon carbide composite
cladding
Conceptual design

Design status

1. Introduction
General Atomics (GA) is developing a 50-megawatt electric (MWe) fast modular reactor (FMR) that provides
safe, carbon free electricity, capable of incremental capacity additions. A modular design allows it to be factorybuilt and assembled on-site to keep the cost of capital low, while the use of dry cooling facilitates siting tocomplement renewables in nearly any location. The FMR is being developed for a demonstration before 2030
and deployment by the mid-2030s.

2. Target Application
The FMR is being developed for the electricity generation with fast response to grid operators and a potential
for high temperature use.

3. Design Philosophy
The FMR is designed to provide flexible and dispatchable carbon-free power for the 2035 US energy market.
The design will develop and verify that simplified characteristics, such as the inert helium gas coolant,pellet-loaded fuel rod, installations free of heat sink requirements, and small and passive heat removal systems,
will result in a safe, maintainable, cost-effective, distributed, nuclear energy-generating station.

4. Main Design Features
The FMR is a helium-cooled fast reactor with a core outlet temperature of 800°C and a grid-capable powersource with a net electrical output of approximately 50 MW.

193

Power Conversion
The direct helium Brayton cycle enables fast grid response and high overall efficiency during normal operation.
The high-speed generator, along with a permanent magnet rotor and the helium turbo-compressor, has a small
rotational inertia. The high-efficiency, solid-state converter accepts a wide range of input frequency andvoltage and outputs the line frequency and voltage with equivalent power.

Reactor Core
The reactor has a compact annular core surrounded by advanced fast neutron reflector material, i.e., zirconium
silicide (Zr3Si2), graphite, and steel vessel. Low core power density and use of uranium dioxide (UO2) fuel andsilicon carbide (SiC) composite matrix ceramic cladding secure the safety margin of the fuel. A three-batchrefueling scheme is used to enhance higher fuel utilization and reduce the requirement of excess reactivity
control.

Fuel Characteristics
The FMR uses conventional UO2 as the fuel material. The average discharge burnup of ~100 gigawatt days
per ton of uranium (GWd/tU) is achieved by use of high assay low enriched uranium (HALEU) andirradiation-resistant SiC composite cladding that is derisked in the current Accident Tolerant Fuel (ATF)
program.

Fuel Handling System
All fuel handling activities are carried out from the floor of the maintenance hall. All access to and removal of
the fuel will be by remote handling. The refueling equipment includes a dry spent fuel storage canister (SFSC)
and a fuel-handling machine.

Reactivity Control
The excess reactivity is controlled by control rods made of boron carbide (B4C) encapsulated in SiC sheath.
Each control rod is fail-safe, by design. The shutdown rod is also made of B4C which is fully withdrawn fromthe core during normal operation.

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) contains the nuclear fuel and reactor internals. The RPV is made of 316
Stainless Steel. The reactor internals include the high-temperature upper support structure, the lower core
support structure, and the in-core instrumentation support structure.

Reactor Coolant System
The reactor coolant system (RCS) removes heat from the reactor core and transfers it to the power conversion
unit (PCU) for electricity generation. The RCS provides controlled pressurized helium for transport of heatfrom the reactor to the PCU during normal operation and from the reactor to the maintenance cooling system(MCS) during shutdown conditions when the PCU is not available.

5. Safety Features
The FMR uses a non-hazardous helium coolant, a chemically inert gas that is non-explosive, non-corrosive,and not activated. The reactor has an annular core surrounded by advanced reflector material and a steel vessel.
Low core power density and use of high-temperature SiC composite cladding secure the safety margin in thehighly unlikely event of an accident. Under such accident condition, the reactor can be cooled over the
long-term by a gravity-driven reactor vessel cooling system (RVCS).

Engineered Safety System Approach and Configuration
The engineered safety features (ESF) are implemented to mitigate the consequences of postulated accidents.
The ESF include reactivity control system, reactor protection system, maintenance cooling system, reactor
vessel cooling system, and containment.

Decay Heat Removal System/Reactor Cooling Philosophy
The normal decay heat removal is conducted by the PCU. The MCS is an active system, including a recirculator
and a heat exchanger. The RVCS provides a redundant, passive means of transporting core residual heat whenneither the PCU nor the MCS is available, preventing the reactor vessel from exceeding design temperature
limits. The RVCS design includes an array of cooling pipes, forming a “cooling panel” placed around thereactor vessel. Heat is removed from the cooling panel by natural circulation of water through the coolingpanel.

Spent Fuel Cooling Safety Approach/System
The spent fuel storage facility (SFSF) stores 60 years of spent fuel. The SFSF provides adequate passive drycooling of the spent fuel, protection from external threats, and monitoring for SFSC leakage. The SFSF is
located below grade and is protected by an extension of the protective shield that covers the reactor building
maintenance hall. The SFSF has redundant, elevated air intakes, and outlets for cooling.
194

Containment System
The containment system uses a below-grade, free-standing, steel containment with concrete backing to enclose
the primary coolant system. The containment is safety-related and is classified as Seismic Category I. The
containment will be supported from a seismic isolation platform and be hermetically sealed during normal
operation.

Chemical Control
The chemical and volume control system (CVS) provides reactor coolant purification, reactor coolant
inventory control and makeup, chemical control, and oxygen control. The CVS performs these functions during
normal modes of operation including power generation and shutdown.

6. Plant Safety and Operational Performances
The FMR combines the safety features of the high-temperature gas reactor (HTGR) with the high fuelutilization feature of the GFR, while taking advantage of UO2 fuel and SiC composite cladding developments
under the ATF program. The high-temperature, radiation resistance of SiC composite cladding enables thepassive safety of the FMR system driven by the RVCS, which does not require bulk heat reservoir or injectionof emergency cooling. The direct helium Brayton cycle enables fast grid response for load following, and highoverall efficiency during normal operation. The automatic control of the reactor power and turbomachinery
keep the reactor at a constant temperature that mitigates thermal cycle fatigue associated with mostload-following reactors. Millisecond response to the grid is provided by the combination of the integrated
direct Brayton cycle reactor and turbomachinery, the permanent magnet (PM) generator, and the frequency
converter. The highly automated, hierarchical control system reduces the burden on the plant operator andallows for efficient grid operator dispatch.

7. Instrumentation and Control Systems
The instrumentation and control (I&C) safety system is designed to achieve an innovative, architecture-like,field programmable gate array (FPGA) technology that promotes simplicity of design and enhances
verifiability of the reactor protection system (RPS). The non-safety system is a distributed control system(DCS) that is a highly automated, hierarchical control system that requires minimum operator actions toperform predefined functions (i.e., startup, normal power-range load-following, turbine warm-up, reactor
shutdown, etc.) as well as during certain off-normal and abnormal operations. The DCS utilizes digitaltechnology, such as microprocessors, FPGAs, and application-specific integrated circuits (ASICs), to takeadvantage of its high reliability and inherent capabilities regarding continuous self-diagnostics, on-line
monitoring, high reliability, real-time data analysis and trending, etc. These elements build a nearlyautonomous system that will enhance a significant reduction in operation and maintenance (O&M) cost.

8. Plant Layout Arrangement
The FMR prototype plant consists of a single reactor module sited below grade. The plant includes a reactor
and support facilities within a minimized footprint of 38,000 m2. Dry cooling structures are located adjacentto the reactor. Support equipment for helium and water processing, cooling, fire suppression, and backup powersystems are all contained within the auxiliary building. All water, helium, and fire suppression systems areprotected in an underground fluid transport network structure connected to the reactor.
.

Compact GA 50 MWe FMR Plant

195

9. Testing Conducted for Design Verification and Validation
The FMR design verification and validation plan will be developed. The FMR fuel rod integrity will be
experimentally verified in an advanced test reactor (ATR) and transient reactor test (TREAT) facility forhigh-burnup irradiation and power ramping, respectively. The high-burnup irradiation tests use the fission
accelerated steady-state testing (FAST) approach to accelerate the irradiation, while the transient tests
characterize the fuel cladding mechanical behavior under the initial power ramp. This test will use a modified
version of the dry in-pile fracture test (DRIFT) capsule.

10. Design and Licensing Status
GA is currently executing pre-application licensing engagement with the Nuclear Regulatory Commission(NRC), including the FMR principal design criteria, quality assurance (QA) program plan, pre-probabilistic
risk assessment (PRA), licensing basis event (LBE) selection, and safety classification, following regulatory
guide (RG) 1.23259, RG 1.23360, and Nuclear Energy Institute (NEI) 18-0461. In the longer term, thelicensing approach for the demonstration and commercialization of the FMR concept is to construct andoperate a demonstration unit through two-step 10 CFR Part 50 licensing with prototype testing features.

11. Fuel Cycle Approach
The baseline FMR design is based on once-through fuel cycle with interim storage and direct disposal.
Recycling is feasible through a proliferation-resistant dry process that uses only thermal-mechanical treatmentof oxide fuel without any involvement of chemicals.

12. Waste Management and Disposal Plan
The radioactive waste management systems segregate, collect, process, store, and dispose of radioactive liquid,gaseous, and solid wastes. The liquid waste system includes filtration and demineralization, discharging
processed wastes to the solid waste system. The gaseous waste system reduces the activity of the waste gas toa level acceptable for discharge to the atmosphere. The solid waste systems include collection, solidification,
packaging, and storage facilities for radioactive materials prior to their offsite shipment for disposal.

13. Development Milestones
2024
2030
2035

196

Conceptual design of reactor and power plant: pre-application licensing engagement;
economic analysis
Risk reduction: demonstration of reactor operation; demonstration of large components;
fuel qualification
Commercialization: prototype reactor and plant; demonstration of plant operation;
licensingXe-100 (X-energy, LLC, United States ofAmerica)
All content included in this section has been provided by, and is reproduced with permission of X-energy, LLC, United States of America.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

X-energy, LLC, United States ofAmerica
Modular HTGR

Coolant/moderator

Helium/graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

200 / 82.5

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

6.0 / 16.5

Number of fuel assemblies in thecore
Fuel enrichment (%)

220 000 pebbles per reactor

Core Discharge Burnup (GWd/ton)

165

Refuelling Cycle (months)

Online fuel loading

Reactivity control mechanism

Thermal feedback & control rods

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)
RPV height/diameter (m)340m x 385m (4 reactor modules
with 4 turbines)
16.4 / 4.88

RPV weight (metric ton)

274

Seismic Design (SSE)

0.5g

Fuel cycle requirements / Approach

Uranium once through (initially)

Distinguishing features

Online refuelling, core cannot
melt and fuel damage minimized
by design, independentradionuclide barriers, potential foradvanced fuel cycles
Basic design

Design status

Forced helium circulation

260 / 750
UCO TRISO/pebbles

15.5

1. Introduction
The Xe-100 is a pebble bed high-temperature gas-cooled reactor with thermal rating of 200 MW. It features acontinuous refuelling system with low enriched fuel spheres or pebbles of approximately 15.5 wt% entering
the top of the reactor and passing through the core six (6) times to achieve a final average burnup of 165 000
MWd/tHM.

2. Target Application
Process heat applications, desalination, electricity and co-generation.

3. Design Philosophy
A major aim of the Xe-100 design is to improve the economics through system simplification, component
modularization, reduction of construction time and high plant availability.

4. Main Design Features
Reactor CoreThe Xe-100 core comprises ~220 000 graphite pebbles fuel elements each containing ~18 000 UCO TRISO
coated particles. The core is graphite moderated with online refueling capability. The advantage of online197

refuelling is that the core excess reactivity is maintained at below 2% which means that no burnable poisons
are needed to ensure that the reactor reactivity remains within safe shutdown limits at all times. This also
improves the neutron economy of the core and helps the Xe-100 to achieve an average burnup of 165 000
MWd/tHM. At full power approximately 173 fresh pebbles are added daily, and a similar number are also
removed as spent fuel.
The core geometry (i.e. aspect ratio), power density, heavy metal loading and enrichment level have been
optimized to ensure that decay heat can be removed during even the most severe accident scenario such as a
total loss of power along with the loss of the helium heat transfer fluid. During such an event, known as a
Depressurized Loss of Forced Cooling (DLOFC), the decay heat is removed passively through making use of
the thermal characteristics of the core and graphite core support structures.

Fuel Characteristics
TRistructural ISOtropic (TRISO) particles are embedded in a graphite matrix pebble to form the fuel element.
Particles contain coated uranium oxide and carbide (UCO) kernels enriched at 15.5 wt% and are slightly
smaller in diameter (425 µm) than the usual UO2 (500 µm) fuel kernels used in Germany and China. The
optimized moderation ratio (NC/NA) yields a heavy metal loading of around 7 g/pebble. This enables the Xe100, under worst case water ingress scenarios, to be shut down with its reactivity control and shutdown system(RCSS). Moreover, the graphite shell does not melt but sublimes (changes into vapor) at > 3920°C (4200K)
and fuel temperature never exceeds 1100°C during normal operation. Therefore, X-energy does not have tobear the same magnitude of costs related to the pressure vessel, containment building, or safety systems as
those of a traditional nuclear plant.

Fuel Handling System
The fuel handling system (FHS) moves fresh fuel pebbles, upon arrival at the plant, to the reactor where theyremain until the fuel has been fully utilized. The pebbles are then removed from the reactor and transferred tothe spent fuel storage system. The FHS comprises four main subsystems/components: new fuel loading system;
fuel unloading and recirculation system; fuel burnup-measurement system; and spent fuel handling and storage
system.
The FHS is a closed system which allows for 100% accountability of the fuel as it enters and exits the reactor.
Each time the fuel passes through the reactor the burnup is measured to determine the amount of useful fuelavailable. If the fuel is not fully spent, it is recycled through the reactor and remains in the fuel handling system
until spent and is then deposited into a spent fuel cask. These casks are stored onsite for the life of the plant.

Reactivity Control
First and foremost, the reactor relies on a strong negative temperature coefficient to ensure nuclear stability at
all times. For operational reactivity control the reactor has a RCSS comprised of a bank of nine control rods
with B4C as the main control poison. A second bank of nine rods remains in the fully withdrawn position acting
as reserve shutdown system primarily used for maintenance shutdown. The negative temperature coefficientalone will shut the reactor down to a safe shutdown condition without the need for active reactivity control
systems. The control rod and shutdown rods can however individually shut down the reactor in a controlled
shutdown operation. To achieve indefinite shutdown at temperatures of about 100°C for maintenance, bothbanks are inserted. Due to continuous fuelling, a minimum excess reactivity margin can be maintained. Thismargin is functionally selected to allow for start-up when performing load-follow operation (100%-40%100%) and is sufficient to cover the effect of Xeon decay.

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) and internal structures are designed for a 60-year life.

5. Safety Features
The intrinsic safety characteristic of the plant is guaranteed by a relatively low power density of 4.8 MW/m3,
high thermal inertia of the graphitic internals and a strong negative temperature coefficient of reactivity over
the total operational regime of the reactor. Also, the use of qualified UCO TRISO coated particle fuel provides
excellent retention of fission products at the source. The pressure boundary provides a further independent
physical barrier to retain the small amount of fission products that may end up circulating in the helium and ingraphite dust particles. The reactor building venting route also minimizes the release of fission products byventing through filtered release vents.

Engineered Safety System Approach and Configuration
The primary engineered safety systems are designed to be passive. Unintended plant transients are avoided
due to the small excess reactivity resulting from continuous fuelling. The RCSS insertion depth during normal
operation binds around 1.4 niles (1 nile = 1000 pcm), allowing for load-follow operation within the range of
100% - 40% -100%. Any spurious signal that would cause full withdrawal of the RCSS would therefore onlytranslate to a higher temperature and will not cause fuel damage.

Decay Heat Removal/ Reactor Cooling Philosophy
Passive decay heat removal is possible, while the fuel temperature remains below admissible values. The
198

radionuclides remain inside the fuel even throughout extreme upset events. If the active heat removal system
is not available, then the core heat is removed passively through: Conduction between the pebbles and side
reflector; Convection and thermal radiation to the core barrel, RPV; and, Reactor Cavity Cooling System(RCCS). Loss of the RCCS does not result in a safety concern as decay heat can be safely dissipated into thebuilding structures and finally to the environment.

Containment Function
Xe-100 ‘functional containment’ is based on TRISO coated particles serving as the primary barrier toradionuclide release. The fuel element matrix contributes to additional resistance and adsorber surface in
diffusing radionuclides. The helium pressure boundary (HPB) is the secondary independent barrier while thereactor building serves as final barrier. In the event of a break in the HPB a building flap will open, serving to
let the helium escape to atmosphere through a filtered release vent to remove radionuclides.

6. Plant Safety and Operational Performances
The design has the following inherent safety characteristics and design features:
Non-metallic fuel elements – meltdown proof and efficient retention of radionuclides in the TRISO- coatedparticle fuel during normal operation allows for relatively clean helium circuits and plant operations with low
contamination of cooling gas and radioactivity release;
Helium – Chemically and radiologically inert helium is an effective heat transport fluid. Moreover, it does not
influence the neutron balance. Helium allows for very high coolant temperatures;
Graphite core structures – allows for high-temperature operations and provides high thermal inertia to thereactor resulting in slow transient response during a loss of active cooling.

Engineered Safety System Approach and Configuration
The following is credited as safety systems (active and passive):- Coated particle fuel elements;- Reactor protection system (RPS);- Core support structures;- RPV;- Reactor building

Operational transients and accidents(i) Key safety features to limit plant transients:
The RCSS insertion depth during normal operation binds around 1.4 δk-eff. Any spurious signal that would
cause full withdrawal of the RCSS would therefore only translate to a higher temperature that would remain
below an allowable value shown experimentally not to cause any fuel damage. Furthermore, because thereactor core and its internals are mostly graphite, this provides a high thermal inertia that would cause anytransient to be slow-acting.
(ii) Key safety features to avoid core damage:Features include the reactor core with a low power density, which is very robust and has a high thermal capacity
to make the reactor thermally stable during all operational and controlled procedures. Strong negative
temperature coefficients also contribute to the excellent inherent safety characteristics.
(iii) Key safety features to contain core damage:Core meltdown proof – no Core Damage Frequency(iv) Key safety features to reduce or eliminate large offsite release;
Multiple – independent fission product barriers:- Qualified UCO TRISO coated particle fuel provides retention of fission products at the source;- ASME designed pressure boundary provides a further reliable physical barrier to retain the small amount offission products that may end up circulating in the helium and in graphite dust particles;- A filtered and vented reactor building.
(v) Diversity and redundancy:
A series of independent fission product barriers provides redundancy and diversity. Failure of any one
individual barrier will not impact the performance of another neighboring system/barrier.
(vi) Worst accident scenario and release:
The Depressurized Loss of Forced Coolant (DLOFC) is the worst-case accident scenario. This assumes the
RCSS has also failed to insert. Under this scenario no fuel damage will be experienced.

7. Instrumentation and Control Systems
The I&C system consists of three layers: Distributed control system, investment protection system, and reactor
protection system. The human machine interface is configured in such a way that no operator action is required
to ensure safe shutdown of the reactor during all events.

199

8. Plant Layout Arrangement

9. Design and Licensing Status
Conceptual design development and U.S. Nuclear Regulatory Commission pre-licensing phase.

10. Development Milestones
2019
2021
2021
2025

200

Conceptual Design Development Complete
Basic Design Development Complete
Applications submitted to the U.S. Nuclear Regulatory Commission
Start of Construction

SC-HTGR (Framatome Inc., United Statesof America)
All content included in this section has been provided by, and is reproduced with permission of, Framatome Inc., United States of America.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

Framatome Inc., United States of America

Coolant/moderator

Helium/graphite

Thermal/electrical
capacity, MW(t)/MW(e)Primary circulation

625 / 272

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)
Reactivity control
mechanismApproach to safetysystems
Design life (years)

Prismatic block HTGR

Forced circulation
6 / 16325 / 750
UCO TRISO particle fuel in hexagonal graphite
blocksAnnular core, 102 column, 10 blocks per column
14.5 average / 18.5 maximum
165½ of the core replaced every 18 months; 18 days
outageControl rods (gravity insertion) Independent
reserve shutdown system (gravity insertion) large
negative temperature coefficient
Active and Passive
80

Plant footprint (m2)8,000

RPV height/diameter (m)24 / 8.5

RPV weight (metric ton)

880

Seismic Design (SSE)

0.5g w/seismic isolators

Fuel cycle requirements /ApproachDistinguishing features

LEU once-through fuel cycle / reprocessing
options for later consideration.
Coated particle fuel; passive decay heat removal;
passive safety; high temperature process steam;vented reactor building; 400m EPZ (emergency
planning zone); underground construction;
modular construction; available in a range ofstationary (s) and mobile (m) power levels:
315(s), 180(s), 54(s), 10(s) and 7m (MWth)Preliminary design

Design status

1. Introduction
The Framatome reference scale SC-HTGR is a modular, graphite-moderated, helium-cooled, high temperature
reactor with a nominal thermal power of 625 MW(t) and a nominal electric power capability of 272 MW(e).
It produces high temperature steam suitable for numerous applications including industrial process heat andhigh efficiency electricity generation. The SC-HTGR builds on Framatome’s experience of HTGR projects, aswell as on the development and design advances that have taken place in recent years for modular design andconstruction of HTGRs. The overall configuration takes full advantage of the work performed on early modular
HTGR concepts such as the General Atomics MHTGR, the KWU/INTERATOM HTR-MODUL, and
ANTARES Framatome’s very high temperature gas-cooled reactor design concept.

201

2. Target Application
The SC-HTGR produces high temperature steam suitable for numerous applications including industrial
process heat, Hydrogen production, and high efficiency electricity generation. The HTGR steam cycle concept
is extremely flexible. The steam cycle is also well suited to cogeneration of electricity. The SC-HTGR is
configured in a verity of stationary and mobile options and power levels.

3. Design Philosophy
The SC-HTGR is designed around proven thermal neutron spectrum, helium-cooled, graphite moderated
reactor technology, and passive decay heat removal. The key to the safety and economic potential of thistechnology is the proven radionuclides retention capabilities of UCO kernel TRISO coated fuel particles.

4. Main Design Features
Power Conversion
The reference four-module steam cycle concept plant uses the standard Rankine cycle turbine-generator set
per reactor module. A single reference design reactor module produces 625 MW(t). This design is alsoavailable in a variety of lower powered stationary and mobile configurations.

Reactor Core
The reactor inlet and outlet temperatures are 325°C and 750°C, respectively. These temperatures were selectedprimarily to support the desired steam outlet conditions for the target markets. These temperatures also allow
the use of SA-508/533, a standard PWR vessel material, for the primary system vessels without requiring
cladding, separate cooling, or special thermal protection. The active core is 10 blocks high. The reference SCHTGR reactor module design can be scaled from 625 MW(t) to 7 MW(t) per module using the same fuelblocks in scaled arrangements.

Fuel Characteristics
The TRISO coated fuel particle consists of a uranium oxycarbide (UCO) fuel kernel surrounded by multiple
ceramic and pyrolytic graphitic coating layers that provide the primary fission product retention barriers under
all design basis accident conditions. The total fuel inventory includes roughly 10 billion such particles percore. The particles are distributed in graphitic cylindrical compacts. Multiple compacts are contained within
hexagonal nuclear grade graphite fuel blocks. The compacts are stacked in fuel holes drilled into the blocks
each fuel holes is surounded by multiple helium flow channels tranporting the fission heat to the steam
generators.

.
The core cycle length for the reference SC-HTGR is between 420 and 540 effective full-power days. This has
been confirmed for the initial core, using an initial core loading of 10.36 w/o 235U enriched particles with apacking fraction of 0.289 for all fuel elements in the core, and for reloads utilizing half-core replacement with
fuel blocks having a 15.5 w/o 235U enrichment and a packing fraction of 0.279.

Fuel Handling System
The fuelling and refuelling are performed using robotic refueling machine with the primary coolant boundary
intact. Following shutdown, the primary system temperature is rapidly reduced followed by the helium
inventory reduction to slightly sub-atmospheric. Refuelling access is then gained through the control rod drive
penetrations at the top of the reactor vessel. The robotic refuelling operation is then commenced using
predetermined fuel and reflector block movement sequences.

Reactivity Control
The large negative temperature coefficient of the modular SC-HTGR, along with its large thermal margins,provide for an inherent shutdown capability to deal with failures to scram the reactor. In addition, two gravitydriven and diverse and active reactivity control systems provide further confidence of the ability to shut down
the reactor.

Reactor Pressure Vessel and Internals
The Reactor Vessel is part of the Vessel System which is the primary pressure-retaining components andincludes the Cross Vessels and Steam Generator Vessels. The reactor core, reflector elements, core support202

structure, and core restraint devices are installed in the reactor vessel. The reactor core components, together
with elements of the reactor internal components, constitute a graphite assembly that is supported on a graphite
core support pillars and restrained by a metallic core lateral support system. The reactor internal components
consist of the upper core restraint elements, permanent graphite side reflector elements, graphite core support
pillars, metallic core support assembly, and the upper plenum shroud.

Reactor Coolant System
The concept of “Reactor Coolant” as a system is unique in gas cooled reactors. The reactor fission heat istransported to the steam generators via a heat transport medium i.e., pressurized helium gas. In a normal shut
down condition the core heat is either removed by the steam generators or transported to a water-cooled
shutdown cooling system (SCS). In an accident condition the core decay and residual heat is absorbed within
the internal graphite structure; subsequently and passively transferred to and removed by the water-cooled
reactor cavity cooling system (RCCS).

5. Safety Features
Engineered Safety System Approach and Configuration
The primary safety objective of the SC-HTGR design is to limit the dose from accidental releases so that thedose limits established by the U.S. EPA Protective Action Guides are met at an exclusion area boundary (EAB)
which is located a few hundred meters from the reactor. To achieve this safety objective, the design uses thehigh temperature fission product retention capabilities of TRISO-coated fuel particles, graphite moderator, andhelium coolant, along with the passive heat removal capability of a low power-density core and an un-insulated
steel reactor vessel. The SC-HTGR is designed to passively remove decay heat from the core regardless ofwhether the primary coolant is present. The concrete walls surrounding the reactor vessel are covered by the
Reactor Cavity Cooling System panels, which provide natural circulation cooling during both normal operation
and accidents, so there is no need for the system to actuate, change modes, or configuration in the event of an
accident. Moreover, the thermal characteristics of the reactor are such that even if the RCCS were to fail during
an accident, the safety consequences would still be acceptable. No powered safety-related systems and no
operator actions are required to respond to any of the accident scenarios that have been postulated for thevarious modular HTGR concepts, including the SC-HTGR, throughout the modular HTGR licensing history.

Decay Heat Removal System/ Reactor Cooling
Philosophy
The SC-HTGR has three heat removal systems. The two main
cooling loops transfer heat to the secondary circuit during normal
operation. When maintenance is being performed on the main
cooling loops, a separate shutdown cooling system is available. Thissystem uses a separate and independent circulator and heatexchanger located at the base of the reactor vessel. These systems
also provide cooling during refueling and normal shutdown
conditions as well as most Anticipated Events and DBEs.
The Reactor Cavity Cooling System (RCCS) , shown above, is aredundant natural circulation water-cooled system that maintains
acceptable concrete temperatures in the reactor cavity during normal
operation and Anticipated Events, and maintains acceptable fuel,vessel, and concrete temperatures during Design Basis Accidents.
Each independent loop of the safety related RCCS consists of heatcollecting panels in the cavity surrounding the reactor vessel
connected by a natural circulation loop to a water storage tank.

Spent Fuel Cooling Safety Approach / System
The fuel blocks in the SC-HTGR family of modular reactors are stationary in the core and are replaced at aneighteen to twenty-four (18-24) months intervals. For the reference SC-HTGR (625 MW(t) per module), one
half of the of the fuel assemblies (510 assemblies) are replaced at each refueling interval. The refueling
operations takes place by the semi-autonomous refueling machine under remote control of the refuelingmachine operator.

Containment Function
The radionuclides containment function in the SC-HTGR is performed primarily by the TRISO fuel coatings.
The graphite core structures, primary coolant boundary, and reactor building provide supplemental
containment capability. The SC-HTGR reactor building is vented to the atmosphere during a primary systemdepressurization phase. The building provides aircraft crash protection and supplemental fission productretention in the event of aircraft crash or a primary system depressurization accident.

203Chemical Control
Approximately 2-percent of the primary circuit helium is continuously extracted by the helium purification
system (HPS), cooled, filtered, purified, and returned to the primary circuit. The primary circuit helium
chemistry is strictly controlled by the HPS which maintains the purity and chemistry of the helium within aprespecified standard limits.

6. Plant Safety and Operational Performances
The nominal electricity generation performance of the SC-HTGR system has been evaluate, considering
preliminary efficiency estimates for the helium circulators, feedwater pumps, turbine, generator, and otherplant electrical loads. In an all-electric mode of operation, the net electrical output from each 625 MW(t)reactor module is 272 MW(e), for a net efficiency of 43.5%.

7. Instrumentation and Control Systems
The SC-HTGR Instrumentation and Controls (I&C) include the instruments used for plant protection,
monitoring, and control. The plant design goal is to utilize commercially proven I&C hardware and softwarewith demonstrated reliability.

8. Plant Layout Arrangement
Each reactor module is located in a separate reactor
building. The standard configuration uses a fully
embedded below grade reactor building design asshown below. This provides structural design
advantages and superior protection from external
hazards. An alternative partially embedded
configuration can be used for sites where a fully
embedded structure is not appropriate. The primaryfunctions of the reactor building are to support the
NSSS primary circuit components and to protect thesystem from external hazards.

9. Testing Conducted for Design Verification and Validation
The Framatome SC-HTGR line of high temperature reactor concept possess high technology readiness levels(TRLs) for most if not all its material needs and components. Therefore, no additional testing and V&V beyond
completion of on-going fuel qualification program (AGR) and Graphite characterization research (AGC)
underway by the U.S.A. Department of Energy (DOE) is necessary.

10. Design and Licensing Status
The SC-HTGR concept design and preparatory work for pre-license application is underway.

11. Fuel Cycle Approach
The high thermal efficiency and high fuel burnup of the SC-HTGR support sustainability for current oncethrough fuel cycles by minimizing spent fuel volume. The HA-LEU once through fuel cycle requires about 6.8
MTHM/GW(e)-yr that equates to a natural uranium feedstock utilization of about 224 MT/GW(e)-yr.

12. Waste Management and Disposal Plan
The storage and disposal requirements will be lessened by about 50% for HTGRs as compared to LWRs. The
spent fuel capacity is 10 years. Coated particles represent excellent radioactivity containment characteristics.
For long term storage or disposal, volume reduction and reprocessing are also options studied in the past.

13. Development Milestones
2022
2026
2027
2027
2028
2033

204

Initial Design – On-going
Basic Design (Pending commitment from FOAK customer)
Start of pre-licencing vendor design review in the USA.
FOAK plant engineering design complete; Secure necessary licenses in the USA.
Start construction of a first full-scale NPP module in the USA.
Commercial operation

PeLUIt / RDE (BRIN, Indonesia)
All content included in this section has been provided by, and is reproduced with permission of BRIN, Indonesia.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginCoolant/moderator

National Research andInnovation Agency (BRIN),
Indonesia
Pebble bed high temperature
gas-cooled reactor
Helium/graphite

Thermal Power

40MWt and 10MWt

Primary circulation

Forced circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

3/6

Reactor type

250 / 750

Number of fuel assemblies in the core

Spherical elements withcoated particle fuel27 000Fuel enrichment (%)

17

Core Discharge Burnup (GWd/ton)80

Refuelling Cycle (months)

On-line refuelling

Reactivity control mechanism
Approach to safety systems

Control rod and small
absorber sphere
Combined active and passive

Design life (years)

40

Plant footprint (m2)

~ 24 000

RPV height/diameter (m)11.1 / 4.2 (inner)

RPV weight (metric ton)

180

Seismic Design (SSE)

0.26g

Fuel cycle requirements / Approach

LEU, open cycle, spent fuelintermediate storage at plant
Inherent safety, no need foroffsite emergency measures
Conceptual design of uprated
40MWt is in progress.
For the initial 10MWt RDE,
site license issued in 2017.

Distinguishing features
Design status

1. IntroductionPeLUIt stands for Pembangkit Listrik dan Uap-panas Industri (Indonesian) that means Nuclear Power Plant
for Cogeneration of Electricity and Industrial Heat. In Indonesia the word also means a ‘whistle’ which
represent the spirit and motivation to start the nuclear power plant era in Indonesia. Its initial name was Reaktor
Daya Eksperimental (RDE). RDE was one of the national programmes to support the National Medium Term
Development Plant (RPJMN) for 2015-2019. The main goal of the RDE development Programme is to build
the national capability to be a nuclear reactor technology developer by mastering the design, construction
project management, commissioning and operation of a nuclear power reactor. Furthermore, the nuclear
reactor type selected for the programme should become the prototype design to be scaled up andcommercialized to contribute in enhancing the national energy supply security. The Pebble Bed Reactor (PBR)
type of High Temperature Gas-cooled Reactor (HTGR) was selected as the technology for the RDE
Programme. A sound safety feature and flexible applications of the PBR technology are among the reasons of
this decision. PBR has a high safety level shown by a small radioactive release to the environment on any
probable accident.
As part of licensing procedure, BATAN (currently BRIN) already received the RDE Site Licensing from205

National Nuclear Regulatory Body (BAPETEN) in January 2017. Design Approval phase was already started
in 2018 and 2019, in which BATAN needs to submit the detail design and safety analysis report of RDE to
BAPETEN. However, due to change in main policy of nuclear reactor development, the licensing processes
was halted. Although the research and development of RDE are still continuing, particularly related to safety
analysis and its cogeneration potential. Since the end of 2021, as worldwide trend of hydrogen as an important
key in energy transition is stronger, cogeneration application for hydrogen production becoming the main focus.
In particular, a collaboration with an energy state-own company called PERTAMINA was started to develop
a hydrogen production system based on PeLUIt/RDE. Guided by the initial techno-economic analysis for thehydrogen production using HEEP Software from IAEA an uprated power level to have a more superheatedsteam and electrical output in parallel is needed. Also based on initial safety analysis, an uprate up to 40MWt,
is possible with generally the same core geometry of previous RDE design while maintaining its passive safety
features.
The general scheme of the PeLUIt system is shown in the figure below. The main components of the nuclear
island system are the reactor pressure vessel (RPV) and its internal, coaxial hot gas duct, main steam blower,
and the steam generator pressure vessel. The rest of balance-of-plant system is common to other power plantsuch as coal power plant. While parts of the RPV-internals are shown in the right side of the figure below suchas the core and the graphite reflector surrounding the core. While uprating the power level to 40MWt, it istargeted that the core and RPV dimension are kept.

PeLUIt/RDE Schematic system (left) and main component of the RPV Internals.

2. Target Application
The initial target of PeLUIt/RDE was as an experimental power reactor for research, which basically focuses
to demonstrate the operational and safety performance of a pebble bed high temperature gas-cooled reactor(PB-HTGR). It will demonstrate the capability of the reactor to survive even in the most severe accident
scenarios such as depressurized loss of forced cooling (DLOFC), in addition also to perform a cogeneration
experiment utilizing its high steam temperature output.
In collaboration with national industries such as PT. Pembangunan Perumahan (PT. PP) and PT. Rekadaya
Elektrika Consult (Reconsult), a direct utilization of PeLUIt as electric source for many islands in Indonesia,
replacing an expensive diesel fuel, was foreseen. A joint cooperation with national industries to study anddevelop the techno-economic and detail design of the reactor was already started in 2018-2019, which was
halted. The new focus is to utilize PeLUIt as a green hydrogen production in collaboration with PT.
PERTAMINA.

3. Design Philosophy
The PeLUIt/RDE is designed based on an established pebble-bed reactor principle. The design employs the
TRISO-based fuel which provide a sound fission product retention capability resulting in allowable release ofradioactive material to the environment in any condition of the core including the most severe postulated
accident. The core and reflector are dominantly composed by a graphite which gives a good heat transfer andneutronic features. High heat conductivity and capacity of the graphite improve the heat transfer characteristicof the design. While it helps to improve the thermal neutron spectrum of in the core due to its effective neutron
thermalization capability. An inert He gas as the coolant avoids any chemical or physical reactions. The nuclear
island is applying a typical side by side arrangement of the RPV and Steam Generator Vessel which connectedby a hot gas duct vessel. Each reactor module includes a reactor pressure vessel (RPV); graphite, carbon, andmetallic reactor internals; a steam generator; and a main helium blower. The thermal power of each reactor
206

module is uprated to 40MWt as an additional to 10 MWt option. Both of this power level option have a different
pebble fuel consumption. The helium temperatures at the reactor core inlet/ outlet are 250/700°C, and steam
parameters is 6 MPa/520°C at the steam turbine entrance. The nuclear island is coupled with a ~13MWe or 3
MW(e) steam turbine.

4. Main Design Features
Reactor Core and Power Conversion Unit
The primary helium coolant works at the pressure of 3.0 MPa. Helium coolant enters the reactor in the bottom
area inside the pressure vessel with an inlet temperature of 250ºC. Helium coolant flows upward in the side
reflector channels to the top reflector and top helium plenum and flow into the pebble bed in a downward flow
pattern. Bypass flows are introduced into the fuel discharge tubes to cool the fuel elements there and into thecontrol rod channels for control rods cooling. Helium is heated up in the active reactor core and then is mixed
to the average outlet temperature of 700ºC and then flows to the steam generator. The hot helium then transfers
its energy to the 243ºC feed water in the steam generator to have a superheated steam of 520ºC at 6 MPa
flowing to the turbine to generate a ~13MWe or ~3MWe.

Fuel Characteristics
Fuel elements are spherical ones. Every fuel element contains 5 g of heavy metal. The equilibrium core has
17% enrichment of 235U. Uranium kernels of ~0.5 mm in diameter is coated by three layers of pyro-carbon and
one layer of silicon carbon. Coated fuel particles are dispersed in matrix graphite with 5 cm in diameter.
Surrounding the fuel containing graphite matrix is a 5 mm thick graphite layer.

Fuel Handling System
The operation mode of PeLUIt/RDE adopts continuous fuel loading and discharging: the fuel elements arepneumatically lifted into the upper part of the reactor, drop into the reactor core using a single fuel loading
tube, then move downward across the core and through a discharging tube at the core bottom. The fuels willpass one-by-one through the singulator. The geometry of discharged fuel elements is checked in the fail-fuelseparator. Failed fuel with geometrical defects will be separated and diverted into the failed fuel cask, while
the good ones will continue to the burn-up measurement facility. Fuel pebbles that already reached the burnup
target will be collected in the spent fuel cask while the other will be redirected back into the core. In average,
a single pebble fuel will pass the core 5 times to reach the average discharge burnup target of 80 MWd/kg. A
Once-Through-Then-Out (OTTO) fuel management scheme is being investigated to have a simpler design
which finally improve its green hydrogen techno-economic performance.

Reactor Pressure Vessel and Internals
The primary pressure envelope of RDE/Micro-PeLUIt consists of the reactor pressure vessel (RPV), the steam
generator pressure vessel (SGPV) and the hot gas duct pressure vessel (HGDPV), which are housed in a
concrete shielding cavity. The material for the RPV is selected based on ASME Section III. The RPV consistsof vessel portion, closure head and nozzles. The RPV internals including the ceramic internal and metallic
internal, also the control rod and control rod drive mechanism. The metallic internal include the core barrel
with guides and supports, lower structure with bottom plate and the top thermal shield. The ceramic internal
include all the bottom, side, and top reflector also the outer carbon brick layer.

5. Safety Features
The RDE/Micro-PeLUIt incorporate the established safety features of PB-HTGR design as follows: (1)Maximum temperature of the fuel is below 1400°C in any condition, even in the most hypothetical accident;(2) With that maximum temperature, the TRISO-based fuels contain all fission products to ensure a nonhazardous released to environment. Passive control safety features by the low power density, a large negative
temperature coefficient, low excess reactivity (due to on-line refuelling). Passive cooling safety features
supported by the physical properties of the graphite which is the dominant material in the core also by the low
diameter design of the core.

(a) Engineered Safety System Approach and Configuration

The RDE/Micro-PeLUIt employ a standard engineering safety system of the PB-HTGR. This engineered
safety system function to localize, control, mitigate and terminate accidents and to maintain radiation exposurelevels to the public below applicable limits. It applied the principle of redundancy, high reliability, diversity,and single failure principle. It includes the ventilated low-pressure containment, reactor cavity cooling system,
safety shutdown and protection systems, primary loop isolation, secondary loop isolation, emergency steam
generator drainage system, and main control room habitability system.

(b) Reactivity control
RDE is equipped with two (2) independent reactivity control or shutdown system, a control rod system and asmall ball shutdown system. Control rods are used for shutdown, fine temperature adjustment and trimming.
Each control rod can move in the side reflector columns independently. The small ball shutdown system isprovided for cold and long-term shutdowns. The small ball shutdown elements are stored above the top thermal
207shield and fall under gravity into reflector columns (slotted holes) by demand. The passive control capabilityof the reactor is supported by its strong negative reactivity feedback and a low excess reactivity.

(c) Reactor Cooling Philosophy

In normal operation, the core is cooled by the helium coolant flowing to the reactor and inlet temperature of
250°C. It reaches 700°C after absorbing the heat from the reactor, then the heat is transferred to thesecondary cycle in the steam generator. Under accident conditions, the cooling of the core depends on its
passive cooling system. After shut-down the core decay heat is dissipated passively through the core
structures to the RPV due to the sound thermal characteristic of the graphite. From the RPV, the heat will betaken by the reactor cavity cooling system (RCCS) which operate passively based on natural circulation.
However, the main function of the RCCS is to protect the concrete of the cavity. Even if the RCCS fails,
basically the fuel temperature can be maintained below the design limit.

(d) Containment Function

Containment capability of the PeLUIt/RDE design is based on multi-barrier system. The TRISO-based fueldesign, in particular the SiC layer act as the first barrier which maintain almost all of the fission product.
From the previous HTR-fuel test, in particular the HTR-10 fuel design which adopted in RDE, the fuelable to maintain its containment capability under the temperatures of 1620ºC which is not expected for any
accident scenarios. The second barrier is the primary pressure boundary which consists of the pressure
vessels of the primary components. The third barrier is the ventilated low-pressure containment as part ofengineered safety feature of the design.

6. Plant Safety and Operational Performances
The PeLUIt/RDE is not yet constructed so no empirical safety and operational performances. However, it isexpected that due to online refuelling feature a better availability factor can be expected compared with otherpower plants operating in a mode of periodic fuel loading. One of the priorities is to conduct experiments toevaluate plant operational safety performance using the PeLUIt/RDE.

7. Instrumentation and Control Systems
In general, the instrumentation and control system of PeLUIt/RDE is similar to those of normal PWR plant.
The Reactor Protection System (RPS) has the capability to measure important parameter related withthe reactor safety including the neutron fluxes for the intermediate and power ranges, high and low Heliumgas temperature, mass flow of the main helium coolant and feed water in the secondary system, and pressurein the primary and secondary system.

8. Design and Licensing Status
For the updated 40MWt PeLUIt/RDE including the hydrogen production system is still in theconceptual design phase. While for the initial 10MWt RDE, the Site License was already issued by the
Nuclear Energy Regulatory Agency (BAPETEN) in 2017. Although initial processes of Design Approval
was already started in 2018 and 2019, the licensing of 10MWt RDE was halted.

9. Fuel Cycle Approach
In the adopted fuel cycle, the pebble fuels that already reach its burnup limit are collected in the spent fuelcasks. These spent fuel casks are placed in the spent fuel storage room in the reactor building. The spent fuelstorage room is equipped with the radiation and temperature monitoring and control system. The casks andthe room are designed to avoid criticality of the spent fuels. The cask is designed so that it can be placedin a standard LWR transport cask and be transported if necessary.

10. Development Milestones
2015-20162017-20182018-20192020-2021
2021-2022

208

Development of 10 MWt RDE Conceptual Design and start of its Site Licensing Phase
Site License issued by BAPETEN; RDE Basic Design development started
Review of 10MWt RDE Design and initial Design Approval phase was performed.
Licensing process of 10MWt RDE was halted. Development of safety analysis is still
progressing.
An uprated 40MWt PeLUIt/RDE conceptual design started with focus to develop a greenhydrogen production system.

HTR-10 (Tsinghua University, China)
All content included in this section has been provided by, and is reproduced with permission of Tsinghua University, China.

MAJOR TECHNICAL PARAMETERS
Value
INET, Tsinghua University,
Technology developer, countryof origin
People’s Republic of China
Reactor typePebble bed modular hightemperature gas-cooled test reactor
Coolant/moderator
Helium/graphite
Parameter

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)

10 / 2.5
Forced circulation
3/4
250 / 700
Spherical elements with TRISO
particles fuel (UO2 kernel)
27 000 spherical fuel elements
17Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

80

Reactivity control mechanism
Approach to safety systems

Control rod insertion/ negative
temperature feedback
Combined active and passive

On-line refuelling

Design life (years)

20 (test reactor)

Plant footprint (m2)100x130

RPV height/diameter (m)11.1 / 4.2

RPV weight (metric ton)

167

Seismic Design (SSE)

3.3 m/s2

Fuel cycle requirements /Approach

17% enriched LEU is needed forsuch a small test reactor; Small
amount of material to be includedin national programme.
To verify and demonstrate thetechnical and safety features; andto establish an experimental basefor process heat applications
Operational

Distinguishing features

Design status

1. Introduction
In 1992, the China Central Government approved the construction of the 10 MW(t) pebble bed hightemperature gas cooled test reactor (HTR-10) in Tsinghua University’s Institute of Nuclear and New Energy
Technology (INET). In 2003, the HTR-10 reached its full power operation. Afterwards, INET conducted many
experiments using the HTR-10 to verify crucial inherent safety features of modular HTRs, including (i) lossof off-site power without scram; (ii) main helium blower shutdown without scram; (iii) withdrawal of control
rod without scram; and (iv) Helium blower trip without closing outlet cut-off valve. The second step of HTR
development in China began in 2001 when the high-temperature gas-cooled reactor pebble-bed module (HTRPM) project was launched.

2. Target Application
The HTR-10 is a major project on the energy sector within the Chinese National High Technology Programme,
serving as the first major step of the development of modular HTGR in China. Its main objectives are to: (1)209explorethetechnology
inthedesign,construction
andoperationof
HTGRs;(2) establish an irradiation and experimental facility; (3) demonstrate the inherent safety features of modular
HTGR; (4) test electricity and heat co-generation and closed cycle gas turbine technology; and (5) performresearch and development work on nuclear process heat application. The aims of this project are to demonstrate
the inherent safety features of the HTGR modular design and test the technologies of electricity generation,
district heating as well as process heat application with modular HTGR.

3. Design Philosophy
The primary pressure boundary consists of reactor pressure vessel, steam generator pressure vessel and hot gas
duct pressure vessel which connects the above two vessels. This arrangement can make the maintenance andinspection of the components easier and mitigate the accident result of water ingress into reactor core if
the steam generator heat transfer tubes might fail.

4. Main Design Features
Reactor Core
The reactor core volume is 5m3, 1.8 m in diameter and themean height is 1.97 m. About 27 000 spherical fuel elementswith 60 mm in diameter will be filled up in the reactor core,
the enrichment of fuel is 17% and the mean discharge burnup is designed to be 80 000 MWd/tU. The reactor core isentirely constructed by graphite materials, no metallic
components are used in the region of the core. At the funnelbottom of the reactor core, there is a fuel-element discharging
tube with a diameter of 500 mm and a length of 3.3 m. At the
tube end the special fuel discharge facility singularise the fuelto be unloaded through a 65 mm diameter pipe that penetrates
the reactor pressure vessel.

Reactor Pressure Vessel and Internals
The pressure vessel unit consists of the reactor pressure vessel, the steam pressure vessel and the hot gas ductpressure vessel. The upper part of the reactor pressure vessel is a cover which is connected via eighty bolts,
and its lower part is a cylindrical shelf with a lower closure head. A metallic O-ring and an Ω-ring are used forsealing between the upper and lower parts. The tube nozzle for irradiation channels and the control rods driving
system are mounted on the cover.

Reactor Coolant System
Cold helium channels are designed within the side reflector for the helium primary coolant to flow upward
after entering the reactor pressure vessel from the annular space between the connecting vessel and the hot gas
duct. Helium flow reverses at the top of reactor core to enter the pebble bed, so that a downward flow patterntakes place. After being heated in the pebble bed, helium then enters into a hot gas mixing chamber in thebottom reflector, and from there it flows through the hot gas duct and then on to the heat exchanging
components.

Steam Generator
The steam generator (SG) is a once through, modular helical tube type. Hot helium from the hot gas duct flows
through its central tube to the top part of the SG and then is fed in above the SG heat transfer tubes. Whileflowing around the tubes, the helium releases its heat to the water/steam side, thereby cooling down from
700°C to 250°C. The cold helium flow is then deflected to the inlet of the helium blower and returns to thereactor along the wall of the pressure vessel. The water flows through the helical tubes from the bottom to thetop. The feed water temperature is 104°C and the steam temperature at the turbine inlet is 435°C. The SG
mainly consists of the pressure vessel, the steam generator tube bundle modules and the internals.

Helium Circulator
The helium circulator is a key component for high temperature helium cooled reactors and therefore an
important component to develop and test in the HTR-10. The helium circulator assures the thermal energy
transfer from the reactor core to the steam generator and operates at 3.0 MPa and at 250°C. The circulator is
integrated into the steam generator vessel and installed on top of the SG. The helium circulator was designed
and manufactured by INET at Tsinghua University and the Shanghai Blower Works Co., Ltd.

Fuel Characteristics
The fuel elements are the spherical type fuel elements, 6 cm in diameter with coated particles. The reactor
equilibrium core contains about 27 000 fuel elements forming a pebble bed that is 180 cm in diameter and 197
cm in average height. The spherical fuel elements move through the reactor core in a multi-pass pattern.

210

Fuel Handling System
The HTR-10 is designed to use spherical fuel elements. Its Fuel Handling System (FHS) is different from therefuelling machines of reactors using rod shaped or block shaped fuel elements. The main feature of the FHS
is to charge, circulate and discharge fuel elements in the course of the reactor operation, or in other words online. For the initial core loading, dummy balls (graphite balls without nuclear fuel) were firstly placed into thedischarge tube and the bottom cone region of the reactor core. Then, a mixture of fuel balls and dummy balls
were loaded gradually to approach first criticality. The percentages of fuel balls and dummy balls were 57%
and 43% respectively. After the first criticality was reached, mixed balls of the same ratio were further loaded
to fill the core in order to make the reactor capable of being operated at full power. The full core (including
the cone region) is estimated to have a volume of 5 m3.

Reactivity Control
There are two reactor shutdown systems, one control rod system and one small absorber ball system. They are
all designed in the side reflector. Both systems are able to bring the reactor to cold shutdown conditions. Since
the reactor has strong negative temperature coefficients and decay heat removal does not require any
circulation of the helium coolant, the turn-off of the helium circulator can also shut down the reactor frompower operating conditions. There are ten control rods placed in the side reflector. Boron carbide (B4C) is usedas the neutron absorber. Each control rod contains five B4C ring segments which are housed in the area between
an inner and an outer sleeve of stainless steel. These are then connected together by metallic joints. The innerand outer diameter of the B4C ring is 60 mm and 105 mm respectively, while the length of each ring segment
is 487 mm. There are 7 holes in the side reflector of the HTR-10 for small absorber ball system.

5. Safety Features
HTR-10 has inherent safety features common to the new generation of advanced reactors, i.e. the reactor
automatically shuts down because of the negative temperature reactivity coefficients and the decay heat is
passively removed from the reactor to the environment. HTR-10 is a new generation reactor whose design isbased on the ideas of module reactors.

Reactivity control
The on-line refuelling leads to a small excess of reactivity, the overall temperature coefficient of reactivity is
negative, and two independent shutdown systems are available.

Decay Heat Removal System
After shutdown, the decay heat will be dispersed from the core to outside of reactor pressure vessel via
conduction, convection and radiation, even in the case of depressurized accident condition. Then the decay
heat can be carried out by two independent trains of passive decay heat removal systems to environment. Two
independent reactor cavity coolers are located at the surface of the reactor cavity. During an accident, the decay
heat is removed to the environment by the passive heat transfer mechanisms, i.e. heat conduction, naturalconvection and thermal radiation.

Containment Function
There are three barriers to the release of fission products to the environment, i.e. the coating layers of the
TRISO coated fuel particles, the pressure boundary of the primary loop and the confinement. In any accidents
the maximum temperature of the fuel elements could not exceed the temperature limit and a significant
radioactivity release can be excluded. In addition, the low free uranium content of fuel elements, the retentionof radioactivity by graphite matrix of fuel elements, and the negligible activated corrosion products in the
primary coolant system will maintain the radioactivity of the primary coolant system at a very low level. In
the depressurization accidents of the primary coolant, the impact of radioactivity release on the environment
will be insignificant. Therefore, it is not necessary to provide containment for the HTR-10. Therefore, aconfinement without requirement of pressure-tightness is adopted.

6. Plant Safety and Operational Performances
There are two operational phases for the HTR-10. In the first phase, the plant is operated at a core outlet
temperature of 700°C and inlet of 250°C. The secondary circuit include a steam turbine cycle for electricity
generation with the capability for district heating. The steam generator produce steam at temperature of 440°C
and pressure of 4 MPa to feed a standard turbine-generator unit. In the second phase (not implemented yet),the HTR-10 will be operated with a core outlet temperature of 900°C and an inlet of 300°C. A gas turbine(GT) and steam turbine (ST) combined cycle for electricity generation is in preliminary design. Theintermediate heat exchanger (IHX), with a thermal power of 5 MW, provides high temperature nitrogen gas of
850°C for the GT cycle. There are other options under consideration to operate HTR-10 in higher temperaturemode.

7. Instrumentation and Control Systems
The control system makes use of the distribution control system (DCS). Full digitalized control room andreactor protection system are used in HTR-10.
211

HTR-10 Control room

8. Plant Layout Arrangement
The HTR-10 plant includes the reactor building, a turbine/generator building, two cooling towers and aventilation center and stack. These buildings are arranged and constructed on an area of 100 x 130 m2. The
HTR-10 plant does not contain a leak-tight pressure containing system. The concrete compartments that house
the reactor and the steam generator as well as other parts of the primary pressure boundary are preferably
regarded as confinement.

9. Design and Licensing Status
HTR-10 is operational. The extension of operation license is undergoing.

10. Fuel Cycle Approach
For the test reactor a once through fuel cycle is initially implemented.

11. Waste Management and Disposal Plan
To be included in the national plan of test facilities.

12. Research and Development Plan
From 1986 to 1990, eight (8) research topics for key technologies were defined: (i) a conceptual design andthe supporting reactor physics and thermal fluid design and safety software codes; (ii) manufacturing processof the fuel spheres; (iii) reprocessing technologies for the thorium-uranium cycle; (iv) core internal graphite
structure design and supporting analysis; (v) helium technology establishment, (vi) pressure vessel designs,(vii) the fuel handling design; (viii) development of special materials.
Before the commissioning, the following engineering experiments were conducted: (i) a hot gas duct
performance test; (ii) measurements to establish the mixing efficiency at the core bottom (limit stratificationand heat streaks); (iii) two-phase flow stability tests on the once-through steam generator; (iv) fuel handling
performance test; (v) control rods drive mechanism performance; (vi) V&V of the digital reactor protection
systems; (vii) measurements to confirm the neutron absorption cross-section of the reflector graphite and (viii)
a performance test for the helium circulator.

13. Development Milestones
1992
1995
2000
2001
2003
2018

212

Project approved
Construction beganFirst criticality
HTR-PM Project is launched
Commission date and full power operation
Restart after upgrade of systems; Melt-wire tests to measure temperatures distribution
conducted.

HTTR (JAEA, Japan)
All content included in this section has been provided by, and is reproduced with permission of JAEA, Japan.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin

Reactor type
Coolant/moderator
Thermal power, MW(t)Primary circulation
Primary coolant pressure, MPa
Core Inlet/Outlet Coolant
Temperature, oC
Fuel type/block array
Number of fuel block in core
Fuel enrichment, wt%

Fuel kernelHigh density PyC
SiC

Coated fuelparticle

Low density
PyC

Fuel compact

Fuel assembly

JAEA in cooperation with
MHI, Toshiba, IHI, Hitachi,
Fuji Electric, NFI, Toyo
Tanso, Japan
Prismatic HTGR
Helium / graphite
30
Forced by gas circulators
4395 / 850 (950 max.)

UO2 TRISO ceramic coatedparticle150
3 – 10 (6 avg.)

Average fuel discharged burnup,
GWd/tHM
Refuelling Cycle, days
Reactivity control mechanism

22 (33 max.)
660 EFPD
Control rod insertion

Approach to safety systems
Design lifetime, years

Active~20 (Operation time)

Plant area, m2
RPV height/diameter, mSeismic Design (SSE)Distinguishing features
Status~200m × 300m
13.2 / 5.5> 0.7m/s2 automatic shutdown
Safety demonstration test
In operation

1. Introduction
The High Temperature Engineering Test Reactor (HTTR) is Japan’s first High Temperature Gas-cooled
Reactor (HTGR) established in the Oarai Research and Development Institute of Japan Atomic Energy Agency(JAEA). The HTTR has superior safety features by using coated fuel-particle, graphite moderator, and heliumgas coolant. With the potential of supplying high temperature heat above 900°C, HTGR can be used not only
for power generation but also for process heat in several industrial fields. JAEA conducted long-term hightemperature operation (950°C/50days operation) to demonstrate the capability of high temperature heat supply.
It then conducted a loss of forced cooling (LOFC) test (at 30% power) to demonstrate the inherent safety
feature of HTGR in 2010. The LOFC test simulates the severe accident in which the reactor coolant flow is
reduced to zero and the reactor scram is blocked. The test result shows that the reactor could be shut down andkept in a stable condition without any operation management. JAEA has accumulated useful data for thedevelopment of future commercial HTGR system though the design, construction, and operation of the HTTR.

2. Target Applications
The objectives of HTTR are to: (i) stablish and upgrade the technological basis for the advanced HTGR; (ii)
Perform innovative basic research in the field of high temperature engineering; and (iii) Demonstrate hightemperature heat applications and utilization achieved from nuclear heat.

3. Design Philosophy
Illustrated in the figure below, the reactor building is designed with five levels of three underground floors andtwo upper ground floors. The reactor building is 18.5 m in diameter, 30 m in height. The cylindrically shaped
containment steel vessel contains the reactor pressure vessel, the intermediate heat exchanger, the pressurizedwater cooler and other heat exchangers in the cooling system.
The reactor core is designed to keep all specific safety features within the graphite blocks. The intermediate
heat exchanger is equipped to supply high temperature clean helium gas for process heat application systems.
The instrumentation and control system are designed to allow operations which simulate accidents and213

anticipated operational occurrences. As the HTTR is the first HTGR in Japan and a test reactor with various
purposes, it incorporates specific aspects regarding safety design. JAEA established the safety designprinciples for HTTR in reference to the ‘Guidelines for Safety Design of LWR Power Plants’, but taking into
account the significant safety characteristics of HTGR and corresponding design requirements as a test reactor.
Crane

Reactorpressure
vesselAir cooler

RefuelingmachineServiceareas

Service
areas

Intermediateheat
exchanger

Reactorcontainment
vesselPressurizedwater cooler
Doublepipe
Cutaway view of the HTTR

4. Main Design Features
Reactor Core
The HTTR reactor consists of reactor internals and core components. The reactor internals comprise thegraphite and metallic core support structures and shielding blocks. They support and arrange the corecomponents, such as fuel blocks and replaceable reflector blocks within the reactor pressure vessel (RPV). The
core components are made up of the same prismatic blocks of 360 mm width across the flats and 580 mm inheight, including replaceable reflector blocks, irradiation blocks, control rod guide blocks, and fuel assembly
blocks. The 2.9m in height, 2.3m in diameter core is surrounded by the permanent reflector made of graphite.
The active core region consists of 30 fuel columns and 7 control rod guide columns while the reflector region
contains 9 additional control rod guide columns, 12 replaceable reflector columns, and 3 irradiation columns.

Fuel
The HTTR employs the TRISO (Tri-structural isotropic)-coated fuel particles (CFPs) with UO2 fuel kernel.
There are four layers surrounding the fuel kernel, including a low-density porous pyrolytic carbon (PyC) buffer
layer, followed by a high-density PyC layer, a SiC layer, and an outer high-density PyC layer. Approximately
13-thousand CFPs are fabricated in a graphite matrix of fuel compact. There are 14 fuel compacts in a fuel rod.
Each fuel assemblies contains 31 or 33 fuel rods.
The fabrication of the first-loading fuel for the HTTR started in June 1995. A total of more than 60-thousand
fuel compacts, corresponding to about five-thousand fuel rods, were successfully produced through the fuelkernel, coated fuel particle, and fuel compact processes. The fuel rods were transferred to the reactor buildingof HTTR, where they were inserted into the graphite blocks to form the fuel blocks. In December 1997, 150
fuel assemblies were completely formed and stored in new fuel storage cells.

Reactivity control system
The HTTR contains two reactivity control systems, including a control rod system and a reserve shutdown
system (RSS). The control rod system comprises of 16 pair of control rods made of B4C. Each pair of control
rods can move individually by control rod drive mechanisms located in standpipes at the top head closure of
the RPV. In the event of a scram, the control rods can freely fall into the core by gravity. There are 7 pairs ofcontrol rod in the active core and 9 pairs in the reflector region.
The RSS is located in the standpipes along with the control rod and can be inserted into the third hole of control
rod guide block. The RSS consists of driving mechanism, hopper, guide tube, etc. The hopper contains B4C/C
pellet. When the RSS is activated, the hopper is opened and the B4C/C pellets drop into the reactor by gravity.
The RSS was designed to be able to make the reactor subcritical from any operation condition at a temperature
range from 27°C to 950°C.

214

Cooling systems
The cooling systems of HTTR are composed of a main cooling system (MCS), an auxiliary cooling system(ACS), and a vessel cooling system (VCS). Under a normal condition, the heat of 30 MW from the reactor
core could be removed by the MCS with two loading modes. One is a single loaded operation mode where
30MW thermal from the reactor is cooled by only the
primary pressurized water cooler. Another is a parallel
loaded operation mode, in which 10 MW and 20 MW
thermal are separately removed by the helium-heliumintermediate heat exchanger and the primary pressurizedwater cooler, respectively. The helium-heliumintermediate heat exchanger of HTTR is operated at thehighest temperature in the world.
The ACS consists of the auxiliary heat exchanger, auxiliarygas circulators, and air cooler. The heat transfer capacity of
the ACS is about 3.5 MW. The ACS automatically starts
up when the reactor is scrammed and the MCS is stopped
abnormally. The residual heat of the core can also be
HTTR coolingremoved by the VCS without the activation of ACS.
systems

5. Safety Features
The reactor delivers fully inherent safety due to three enabling design features:- Helium coolant is chemically stable. It does not react chemically with fuel and core structures so thathydrogen gas is not produced by chemical reaction of fuel element in accident like LWR;
- The CFPs of HTTR have excellent heat-resistant property which can bear very high temperature condition
over 2200°C without any fission product release. The HTTR is designed that the fuel temperature does notexceed 1600°C in any accident to prevent fuel damage;- Graphite-moderated reactor core provides a negative reactivity coefficient, low-power density, and highthermal conductivity. Graphite core structure also can withstand up to 2500°C without any thermal damage.
The HTTR can remove the residual heat of the core inherently because of optimized low reactor power density
and graphite core structure. If the forced cooling performance was lost in an accident, decay heat of fueltransfers to reactor vessel through the core graphite structure slowly by thermal conduction and radiation. The
fuel temperature is kept below the design limit of 1600°C by this safety features. The HTTR does not need toconsider the immediate accident management and to provide excess emergency safety system.

6. Plant Safety and Operational Performance
Various operational tests have been conducted to confirm the plant safety and operational:a) Pre-operational test
Pre-operational test operation of the reactor cooling system was performed from May 1996 to March 1998.theAtstage of the pre-operational test without nuclear heating the helium gas was heated by the gas circulators
up toabout 200°C at 2 MPa. Plant control systems were also fully checked. During the pre-operational tests,
severalimprovements in the system were made in terms of securing its safety margin and easy operation. Theirperformance was finally confirmed in July 1999 after completing the actual fuel loading.
b) Start-up physics test
Fuel loading to the reactor started in July 1998, and the first criticality was attained on November 10th, 1998.
The fuel blocks were column-wise loaded from the outer fuel columns to the inner. The first criticality wasachieved successfully with 19 fuel columns loading. After that, the other inner fuel columns were loaded andthefull core criticality was achieved by December 1998. In the course of fuel loading, low power physics tests
werealso carried out for the 21, 24, and 27 fuel columns loaded core. These tests provide useful data fordesigning future annular cores of advanced HTGR.
c) Rise to power test
Rise to power tests were started in September 1999 when the reactor power was increased step-by-step to 10
MW(t), 20MW(t), and then finally to 30 MW(t). The 30 MW(t) full power and 850°C high reactor outlet
coolant temperature were achieved in December 2001. Certificate of pre-operation test, that is, operation
permit of the HTTR was issued in March 2002. The HTTR accomplished the maximum reactor outlet coolant
temperature of 950°C in April 2004 in high temperature test operation. Operation permit for the hightemperature test operation was issued in June 2004.
d) Safety demonstration test
The safety demonstration tests have been implemented from 2002 in order to confirm the excellent inherent
safety of HTTR. In the first phase of the safety demonstration test, the control-rod withdrawal tests, the gas
circulator tripping tests, etc. have been carried out demonstrating the safety of HTTR. From 2010, the second
phaseof safety demonstration tests, namely loss of forced cooling test (LOFC), was carried out. The LOFC
test wasinitiated by tripping all three helium gas circulators of the HTTR while deactivating all reactorreactivity
control systems to disallow reactor scram due to abnormal reduction of the primary coolant flow
rate. The test
results showed that the reactor power immediately decreased to almost zero and became stable215

as soon as the helium gas circulators were stopped as shown in the figure below.

e) Continuous operation
In order to demonstrate the long-term operation of the heat utilization system, the HTTR was conducted
Rated/Parallel-loaded 30 days, and further 50 days, continuous operation with full power. This was the firstlong-term operation with a reactor outlet coolant temperature over 900°C. The continuous operation test
confirmed that the reactor internal structures and the intermediate heat exchanger, which are the coretechnologies of the HTGR, operated properly as designed value. The intermediated heat exchanger could alsotransfer stable high-temperature heat from the primary to secondary helium coolant.

7. Instrumentation and Control Systems
Instrumentation and control systems consist of instrumentation, control, and safety protection systems. The
instrumentation includes reactor and process instrumentations to provide important parameters such as control
rod position, neutron flux, temperature, pressure, flow rate, etc. for operation, monitoring, and reactor
protection. There are about four-thousand sensors in the HTTR, and the signals from the sensors are centralized
by the plant computer. The control systems comprise the operation mode selector, reactor power control system,and plant control system. The safety protection systems consist of the reactor protection system and engineered
safety features actuating system to ensure the integrity of the core and prevent the fission products release.

8. Plant Layout Arrangement
The plant area is 200 m × 300 m in size, including the reactor building, cooling towers, exhaust stack,laboratory building, and other auxiliary facilities. The reactor building is located in the centre of the plant. The
exhaust stack is on the north side of the reactor building to ventilate the air from the reactor building to theatmosphere. The laboratory building and the development building are on the west of the reactor building.

9. Design and Licensing Status
The HTTR construction started in 1991 with first criticality accomplished in 1998. Details of the operational
achievements is given below. The HTTR has not been in operation since the great east Japan earthquake
occurred in March 2011. In conformity with new regulatory requirements by Nuclear Regulation Authority ofJapan, JAEA received permission and restarted operation of the HTTR on July 2021.

10. Development Milestones
1969 – 1984
1985 – 1990
1991 – 1997
1998
2001, 2002
2004
2007; 2010
2014
2020
2021.07.30

Conceptual design (4 years); System integrity design (6 years) and Basic design (3 years)
Detail design (3 years); Application and permission of construction (1 year)
Construction
First criticality
Reactor outlet coolant temperature of 850οC; Safety demonstration test
Reactor outlet coolant temperature of 950οC
850οC/30 days operation; 950οC/50 days operation and Safety demonstration tests
Conformity review on the New Regulatory Requirements start toward resume operation
Permission toward the restart of HTTR
Operation restarted

11. Future plans
Following the restart of the HTTR, a safety demonstration test in the OECD/NEA LOFC project was carriedout on January 2022. New test project to demonstrate hydrogen production by 2030 using the high temperature
heat from the HTTR has been launched. Other than these, a number of activities are planned to be carried out
through operation of the HTTR including international cooperation and human-resource development. As the
HTTR can be used as a test bed for international cooperation, JAEA plans to launch new international projects
based on the operation of the HTTR, and welcomes discussion with potential partners.

12. Reference
T. Nishihara et al., “Excellent Feature of Japanese HTGR Technologies”, JAEA- Technology 2018-004.

216

PART III. LIQUID METAL COOLEDFAST NEUTRON SPECTRUM
SMALL MODULAR REACTORS

217

218

BREST-OD-300(NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of, NIKIET, Russian Federation.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
NIKIET, Russian Federationof origin
Reactor type
Coolant
Thermal/electrical capacity,
MW(t)/MW(e)Liquid metal cooled fast reactor
Lead
700 / 300

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
0.1 / 17 – 18.5

Core inlet/outlet coolant
temperature (oC)420 / 535

Fuel type/assembly array
Number of fuel assemblies in thecore
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)
Reactivity control mechanism

Mixed uranium plutonium nitride
169

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approachDistinguishing features

Design status

up to 14.5
36 – 78
61.45
Reactivity compensation (RC),emergency protection (EP) andautomatic control (AC) members
Passive30
80 × 80
17.5 / 26
27 000VII-MSK 64
Closed fuel-cycle. It uses nitride of
depleted U with Pu
High level of inherent safety due tonatural properties of the lead, fuel,core and cooling design
Construction in progress

1. Introduction
BREST-OD-300 innovative lead-cooled fast reactor is developed as a pilot demonstration prototype for basetype commercial reactor facilities of the future nuclear power industry with the closed nuclear fuel cycle(NFC). The reactor is fuelled with uranium plutonium mononitride (U-Pu)N and uses a two-circuit heattransport system to deliver heat to a subcritical steam turbine and generate electricity of 300 MW(e). The
design documentation has enabled the approval for the construction of the BREST-OD-300 power unit as partof the pilot and demonstration energy complex (PDEC) aimed at the conformity with current common andspecific standards for lead-cooled reactor facilities. Endurance of the reactor facility will be demonstrated. An
extensive R&D program, required for commercial lead-cooled reactor facilities, will be carried out.

2. Target Application
The BREST-OD-300 reactor is designed for different modes of operation and optimizing all processes andsystems that support the reactor operation. The main goal is practical confirmation of the ‘inherent safety’concept of the lead-cooled fast reactor, operating in NPP mode in NFC. After operational tests, the unit will
be commissioned for electricity supply to the grid.

3. Design Philosophy
BREST-OD-300 is a pool-type reactor design with metal-concrete vessel. The objective of the design is toeliminate severe accidents; complete fuel breeding (equilibrium mode) for self-sustaining and exclusion of219

accidents caused by reactivity; integral-type arrangement of the primary circuit to avoid release of coolant
outside the reactor vessel, to eliminate the loss of coolant; to use of low-activated lead coolant with high enoughboiling temperature, without adverse interaction with water and air in the case of depressurizing of the circuit.
The reactor facility has a two-circuit steam generating power unit that includes reactor core, steam generators(SG), main circulation pumps (MCP), fuel assembly (FA) loading system, control and protection system(CPS), steam-turbine unit, passive decay heat removal system, reactor heat-up system, reactor overpressure
protection system, gas purification system and other auxiliary systems.

4. Main Design Features(a) Power Conversion
The pool-type reactor design has an integral lead circuitaccommodated in one central and four peripheral cavities ofthe metal-concrete vessel. The central cavity houses the core
barrel together with the side reflector, the CPS rods, the spent
fuel assembly (SFA) storage and the shell that partitions the
hot and the cold lead flows. Four peripheral cavities(according to the loop number) accommodate the SG and
MCP, heat exchangers of the emergency and normal cooldown systems, filters of coolant and auxiliary components.
The cavities have hydraulic interconnection.

(b) Reactor Core
The lead coolant properties in combination with dense, highheat-conductivity nitride fuel provide conditions for complete
plutonium breeding in the core (CBR ≥ 1). That results in asmall operating reactivity margin (Δρ < βeff) and enables
power operation without prompt criticality power excursions.
The adopted fuel is mixed mononitride (U-Pu)N that featureshigh density (14.3 g/cm3) and high conductivity (20 W/m·K)
and is compatible with lead and the fuel cladding of chromium
ferritic-martensitic steel. The FA design allows radial coolant
flow transfer in the core which prevents overheating of thedamaged FA as shown in the figure on reactor core map.

BREST-OD-300 Layout(c) Reactivity Control
When operating a core with a "zero" reactivity margin for fuelburnup during a micro campaign, all reactivity compensation
BREST-OD-300 Reactor Core Map(RC) control members and emergency protection (EP) control
members are removed from the core. Only two pairs of automatic control (AC) members remain partiallyinserted, compensating for the neptunium effect of reactivity ~ 0.08 % Δk/k and small reactivity swings whenisotopic composition of fuel in the core changes. There is no need for the group consisting of all RC control
members to participate in compensation of reactivity loss during fuel burnup; the function of RC control
members is to compensate the temperature and power effect of reactivity; they are partially inserted into thecore in the cold state at a minimum power.

(d) Reactor Pressure Vessel and Internals
An integral pool-type layout is used in the reactor facility to avoid coolant losses. The reactor vessel material
is multilayer metal concrete; the lead coolant and the main components of the primary circuit are located in
the reactor vessel. The central cavity accommodates the reactor core with side reflector, the CPS rods, an inreactor SFA storage and a reactor core barrel that separates the hot and cold lead flows. The four peripheral
cavities (one for each loop) accommodate steam generators and reactor coolant pumps, heat exchangers of theemergency and normal cooldown systems, filters and other components. The cavities are hydraulicallyinterconnected.

(e) Reactor Coolant System
Heat is removed from the reactor core through forced lead coolant (LC) circulation by pumps. The LC is
pumped to the height of ∼2 m relative to the lead level in the suction chamber and supplied to the free level ofthe annular pressure chamber. The lead further goes down to the core support grid, flows upward through thecore where it is heated up to the temperature of 535°C, and enters the shared ‘hot’ coolant drain chamber. Thencoolant flows up and enters the SG inlet cavities and inter-tube space via the distributing header nozzles. As
flow goes down into the inter-tube space, the LC transfers heat to the secondary coolant flowing inside the SG
tubes. Cooled down to ~ 420°C, the LC goes up in the annulus and flows out the pump suction chamber, where
it is pumped out back to the pressure chamber. Exclusion of high pressure in the primary lead circuit and a
relatively high lead freezing temperature contribute to crack self-healing, which eliminates the possibility ofloss-of-core-cooling accidents and release of radioactive lead from the reactor vessel. Lead circulation through220the reactor core and steam generator takes place due to the difference between the levels of cold and hot coolant
generated by the pumps. Non-uniformity of lead flow through the steam generators with one of all pumps shut
down is excluded, in so doing flow inertia in fast pump shutdown is provided by equalizing coolant levels indischarge and suction chambers.

(f) Secondary System
The use of chemically inert, high-boiling molten lead in the primary circuit allows adoption of a two-circuit
unit configuration, with a subcritical steam system as secondary circuit. The secondary circuit is a nonradioactive circuit consisting of one turbine unit with subcritical steam parameters, main steam lines, afeedwater system, secondary side of SGs located in the primary circuit. A standard K-300-15, 70-50 turbine
unit with two-cylinder (HPC+LPC) steam condensation turbine with intermediate steam superheating and arotation speed of 3000 rev/min is used. The nominal steam flow rate to the turbine is about 1500 t/h. Oxygen
neutral water at subcritical pressure is used in the secondary loop.

(g) Steam Generator
Steam generator is designed with single-walled twisted tubes, corrosion resistant in water and lead, no welds
along the entire length.

5. Safety Features(a) Engineered Safety System Approach and Configuration
To guarantee the safety the preferential use: neutronic properties and physico-chemical properties of fuel,coolant, materials, as well as special design solutions that allow to fully realize these properties. The leadcoolant properties make it possible to implement in the BREST-OD-300 fast reactor the following: in combination with application of (U-Pu)N fuel, complete breeding of fissile materials in the reactor core,
which provides for a constant small reactivity margin preventing the disastrous effects of an uncontrolled
power increase when introducing the reactivity margin because of equipment failures and personnel errors; to avoid the void effect of reactivity due to a high boiling point and high density of lead; to prevent coolant losses from the circuit in an event of the vessel damage due to high melting/solidification
points of the coolant and the use of an integral (pool-type) layout of the reactor;
 to provide for high heat capacity of the coolant circuit which decreases a possibility of fuel damage; to allow for utilization of the high density of lead and its albedo properties for flattening the FA power
distribution and the fuel pin temperatures respectively, as well as in the safety systems;
 to facilitate larger time lags of the transient processes in the circuit, which makes it possible to lower therequirements to the safety systems’ rate of response.

(b) Decay Heat Removal System / Reactor Cooling Philosophy
The basic principles: no shutoff valves in the primary circuit (no circulation can be lost), a coolant circulation
pattern with a free level difference (circulation is safely continued during loss of power), use of an emergency
cooldown system with natural circulation and removal of heat to the atmospheric air.

(c) Spent Fuel Cooling Safety Approach / System
The emergency core cooling system (ECCS) uses pipes, immersed directly in lead of the primary circuit, which
may be used to cool down reactor under normal conditions. The system coolant circulation in emergency heatremoval mode is provided by natural circulation, with the system coolant under atmospheric pressure. The
system consists of four (4) loops. The ECCS air circuit inlet air temperature operates at a minimum andmaximum temperature of -55oC and 37oC respectively. The system is passive.

(d) Containment System
The localizing function is performed by multilayer metal concrete vessel. Protection from external influences
and threats is provided by the reactor building.

(e) Chemical Control
Regulations for lead coolant including cleaning and decontamination were developed and confirmed by theexperience in operating lead test benches. The absence of corrosion of materials in excess of the specified
limits at a regulated oxygen concentration in lead of (14) ×10-6 was justified.

6. Plant Safety and Operational Performances
An innovative fast reactor BREST-OD-300 with inherent safety is being developed as a pilot anddemonstration prototype for the basic commercial reactor facilities of future nuclear power with a closed
nuclear fuel cycle with a view to the following:- practical confirmation of the key design decisions used for the lead-cooled fast reactor facility operating in aclosed nuclear fuel cycle and of the fundamental guidelines in the inherent safety concept on which these
design decisions are based;- phased justification of the reactor component service life for the creation of commercial nuclear power plants
with lead-cooled fast reactors.
221

The probability of severe accidents due to internal reasons without core destruction involving fuel and cladding
melting, coolant boiling, disruption of circulation in the primary circuit does not exceed 6.48∙10-9 1/year; therelease of radionuclides per day under the conservative scenarios does not exceed the control value. The
absence of the need for evacuation and resettlement of the population in severe accidents at the power unit isensured with a probability of 3.2∙10-8 1/year.

7. Instrumentation and Control System
In the reactor monitored temperatures, coolant levels, oxygen concentration, activity of the lead coolant andcover gas. Control and protection system is based on 2 channel and 3 sets.

8. Plant Layout Arrangement
Plant main building consists of the reactor containment building, auxiliary building, compound building(CPB), emergency diesel generator building and turbine-generator building (TGB). The reactor building ismounted on a single monolithic reinforced concrete foundation plate. In order to reduce seismic inertia forces,
the building is designed to be symmetrical with the footprint of 80×80 m.

9. Testing Conducted for Design Verification and Validation
The complete detailed design of the BREST-OD-300 reactor facility has been carried out. To date,experimental justification of components, elements and equipment of reactor facilities has been carried out
using small- and medium-scale mock-ups and pilot models. Verified and certified software tools were used forcomputational design justification.

10. Design and Licensing Status
The BREST-OD-300 unit design received apositive assessment of the Glavgosexpertiza(2018). Expert review of the Russian Academyof Sciences (RAS) was carried out, which
confirmed that the design of the BREST-OD300 power unit corresponds to the current levelof science and technology, scientific ideas
about the problems of the existing nuclear
power and ways to solve them. The RAS
recommended the construction of the BRESTOD-300 power unit (2019). Following theresults of a long, detailed design review, inFebruary 2021 Rostekhnadzor issued a license
for the construction of the power unit with
Pouring the first concrete into the foundation of the power unit
with the BREST-OD-300 reactor
BREST-OD-300. A solemn ceremony was heldon June 8, 2021, with pouring of the “first concrete” marking the beginning of the power unit construction(Fig. 3). The equipment fabrication and installation and construction of BREST-OD-300 is planned to becompleted in 2026 and begin preparations for the reactor start-up.

11. Fuel Cycle Approach
Establishment of CNFC for full utilization of energy potential of natural raw uranium. Mixed nitride fuel withhigh density and thermal conductivity allows to ensure full reproduction of fuel in the core (core reproduction
ratio ~ 1.05) and compensation of reactivity at fuel burnout. The fuel type considered for the first core and thefirst partial fuel reloads of the BREST-OD-300 fast reactor is nitride of depleted uranium mixed withplutonium, whose composition corresponds to that of irradiated (spent) fuel from VVER’s following
reprocessing and subsequent cooling for ~ 25 years. After completion of the initial stage the reactor operates
in a closed fuel cycle. For the production of fuel, it uses own spent fuel reprocessed and purified from fission
products.

12. Waste Management and Disposal Plan
Progressive approximation to radiation-equivalent (in relation to natural raw materials) RW disposal – at theoperating stage after development of fuel with MA.

13. Development Milestones
1995
2002

2016
2021
2026

222

Conceptual design development initiatedFeasibility study of the BREST-OD-300 NPP with an on-site nuclear fuel cycle(OSNFC)
Design study of the BREST-OD-300 NPP with an on-site nuclear fuel cycle (OSNFC)at the Tomsk Site
Start of construction of the power unit with the BREST-OD-300 reactor
First of a kind pilot demonstration plant starts operation

ARC-100(ARC Clean Energy, Canada)
All content included in this section has been provided by, and is reproduced with permission of ARC Clean Energy, Canada

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country of origin

ARC Clean Energy, Canada

Reactor type

LMFR (pool type)

Coolant

SodiumThermal/electrical capacity,
MW(t)/MW(e)Primary circulation

286 / 100NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant Temperature(oC)
Fuel type/assembly array

Non- pressurized

Forced circulation

355 / 510

Number of fuel assemblies in the core

Metal fuel (U-Zr alloy) based onenriched uranium
99

Fuel enrichment (%)

Avg. 13.1

Refuelling Cycle (months)

240

Core Discharge Burnup (GWd/ton)77

Reactivity control mechanism

Control Rods

Approach to safety systems

Passive, diverse, redundant

Design life (years)

60

Plant footprint (m2)

56 000RPV height/diameter (m)16.7 / 7.9

RPV weight (metric ton)

600

Seismic design (SSE)

0.3 PGA

Fuel Cycle requirements/approach

Metallic HALEU/Open cycle

Distinguishing features

Inherent safety with passive,
diverse and redundant decay heatremoval. Core lifetime of 20 years
without refueling.
Preliminary design

Design status

1. Introduction
The ARC-100 is an advanced SMR that utilizes proven prototype experience while integrating modern design
improvements. It is a 100 MW(e) sodium-cooled, fast flux, pool-type reactor with metallic fuel that builds on
the 30-year successful operation of the EBR-II reactor, built and operated by the Argonne National Laboratory
in the U.S. The ARC-100 effectively addresses the four challenges which have limited the public acceptanceand expansion of the nuclear industry. First, its 100 MW(e) electrical generation capacity is less than one-tenth
the capacity of traditional nuclear power plants, and, consequently, its upfront cost will be affordable by amuch broader range of customers in both the developed and developing worlds. Second, because its coolant isliquid sodium instead of water, its ‘fast’ neutrons have much more energy, giving it the capacity to be fuelledwith and recycle its own used fuel. Third, the ARC-100 utilizes a metallic alloy of uranium instead of uraniumoxide, which provides the foundation for its inherent, walk away safety. And fourth, the operator refuels thispower plant only once every 20 years, rather than every 18 – 24 months which is typical of the light water
reactors which dominate the current worldwide market. The long refueling cycle reduces operational costs andcomplexity, opening markets in the third world and many isolated off-grid applications. Replacement of theentire 20-year fuel cartridge and its removal by the vendor for recycling greatly reduces the risk of nuclear
proliferation.

2. Target Application
The global energy industry is searching for an affordable, flexible, and mature utility-scale nuclear power
solution to address the rapidly evolving energy market landscape and the geopolitics of environmental
regulation. The ARC-100 offers has a unique solution to these challenges – a solution which can deliver:223

Breakthrough economics; Flexible operations and load following to complement intermittent renewable powersources; Technical maturity and demonstrated industrial reliability; Inherent safety performance; The ability
to address the important issue of nuclear waste by recycling over and over its used fuel.
The ARC-100 is a low-risk, low-cost, clean energy solution that is ready for near-term development anddeployment. It will initially target grid-scale electricity generation markets in the developed world. Also, its
inherent safety and simplicity of operation make it ideally suited to satisfy electricity needs at remote locations
like mine sites and the smaller grid markets in the developing world that can accommodate not much morethan 100MW(e). It will also be targeted at industrial heat, hydrogen production and water desalination.
Applications.

3. Design Philosophy
The philosophy of the ARC-100 reactor is to rely on simple, passive safety features to achieve reactor safety
under any normal operational occurrence or accident condition. The ARC-100 has adopted five traditional
levels of safety for its defence in depth: i) Minimize risk by the prevention of abnormal operation and failure
by maximizing safety margins; ii) Protection against abnormal operations and anticipated events via the large
thermal inertia of the sodium pool; iii) Protection against DBA through diverse and redundant systems; iv)
Control of severe plant conditions through designed passive and inherent safety characteristics of the facility;v) Protection of the public health and safety in case of accidents by designing the inherent and passive safetycharacteristics such that operator intervention and external power are not required for plant survival.
Additionally, the design goal of the plant is that the evacuation zone is limited to the site boundary.

4. Main Design Features
Power Conversion
The power conversion system of the ARC-100 consists of a sodium to water superheated steam generator thatpowers an air cooled non-reheat turbine generator set capable of producing 115 MWe.

Reactor Core
The ARC-100 core consists of driver assemblies successively surrounded by steel reflector assemblies andshield assemblies. The core is divided into inner, middle, and outer core zones to flatten the radial power
distribution. The fuel assembly contains fuel pins, each of which provide a plenum to contain fission gases.
The fuel is U-10%Zr binary metallic fuel with an average uranium enrichment of 13.1%. The maximum
enrichment will be in compliance with IAEA requirements.

Reactivity Control
The ARC-100 core employs two independent, safety grade, reactivity control systems. The Primary System
consisting of six control rods is designed to have sufficient reactivity worth to bring the reactor from any
operating condition to cold subcritical with the most reactive control assembly stuck at the full power operating
position. Any operating condition includes an overpower condition together with a reactivity fault. The
Primary System also serves to compensate for the reactivity effects of the fuel burnup and axial growth of themetal fuel. The reactivity associated with uncertainties in criticality and fissile loading is accommodated by
the Primary Control System. The Secondary Control Rod System consisting of three control rods is designed
to shut down the reactor from any operating condition to the cold shutdown, also with the most reactive
assembly inoperative.

Reactor Pressure Vessel and Internals
The Reactor Vessel contains the nuclear fuel and forms the coolant loop of the Primary Heat Transport System.
The primary coolant boundary is completely enclosed within the Reactor Vessel shell and top plate whichforms a pool of sodium. A cover gas of argon, held at pressure slightly higher than atmospheric pressure,resides above the sodium pool within the Reactor Vessel. The Primary System boundary consists of the Reactor
Vessel, the reactor top plate, and the top plate mounted components, principally the intermediate heatexchanger. A Guard Vessel surrounds the Reactor Vessel shell to serve as a leak jacket should the Reactor
Vessel shell develop a leak. The core support structure uses a welded connection at the bottom head of theReactor Vessel. Other than the core support connections and shipping restraints, the vessel has no attachments
and no penetrations below the reactor top plate. Reduced number of penetrations below the reactor top plateand low operating pressure precludes any pipe ruptures. It is a key factor in the ability to keep the corecontinuously cooled for the entire spectrum of design basis events. The design lifetime of the reactor vessel,
as well as the other components, is not less than 60 years.

Reactor Coolant System
The primary circuit of the ARC-100 is the coolant loop of the reactor core which is contained within the reactor
vessel. Two Intermediate Heat Exchangers (IHX) serve as the method of transferring heat to the Intermediate
Heat Transfer System (IHTS), both located within the reactor vessel. The IHX’s penetrate the redan from the
hot sodium pool and transfer heat to the cold pool. Four submersible EM pumps provide forced sodium
circulation within the reactor.

Secondary System
The secondary system is referred to as the Intermediate Heat Transport System (IHTS). The IHTS is the fluid224system for transporting reactor heat between the IHX and the steam generator. It consists of two piping loops
between the IHX, which resides in the Reactor Vessel, and the steam generator. Each piping loop includes an
EM pump and permanent magnet flowmeters in the cold leg. The system also includes instrumentation for
detecting steam generator tube leaks and a rupture disc driven pressure relief line for overpressure protection
for the steam generator shell, intermediate piping and IHXs. The steam generator, the sodium dump valve, theIntermediate Sodium Processing System are located in the Steam Generator Building (SGB).

Steam Generator
The steam generator is a helical coil, single wall tube, vertically oriented sodium-to-water counter-flow shelland-tube exchanger. The steam generator provides the interface for where the sodium flowing in from the
IHTS heats the water to generate superheated steam for the steam turbine plant. Sodium is distributed throughthe shell side of the steam generator while the water flows through the helical coil tube bundles. The steam
generator includes a cover gas space in the upper head of the steam generator which accommodates sodium
level changes due to intermediate sodium thermal expansion and pump transients.

5. Safety Features
Engineered Safety System Approach and Configuration
In the design, the reactor’s inherent reactivity feedbacks have been leveraged for safety and economics in two
ways: (i) The systems provide the basis for the reactor to passively self-regulate its power production to matchthe heat demand from the power conversion cycle without moving control rods (passive load following); (ii)
The systems provide ‘defense-in-depth’ protection to arrest accident progressions before the reactor reaches
unsafe conditions, even if the control and safety rods fail to scram.

Decay Heat Removal System / Reactor Cooling Philosophy
The Balance of Plant steam turbine generator system is relied upon for normal shutdown heat removal. The
ARC-100 emergency Heat Removal systems consist of the following: (i) Direct Reactor Auxiliary Cooling
System (DRACS); (ii) Reactor Vessel Auxiliary (Air) Cooling System (RVACS). The RVACS always
operates in a passive state, RVACS removes the reactor's decay heat through the Reactor Vessel and Guard
Vessel walls by radiation and convection to naturally circulating air outside the Guard Vessel without
exceeding structural temperature limits. The DRACS is composed of three units operating in a naturalconvection mode. Heat exchanger loops using NaK as an intermediate transfer heat from the cold sodium poolto air heat exchangers in which the hot air vents to the atmosphere. DRACS can be either passive or active
using forced air convection

Spent Fuel Cooling Safety Approach / System
Used fuel assemblies are temporarily stored in-vessel at the outside of the core. Once sufficiently cooled, theused fuel assemblies are periodically extracted and placed into an on-site dry-cask storage.

Containment System
The ARC-100 containment system is a Low-Leakage containment type, where the reactor vessel is designed
to operate at near atmospheric pressure. Damage to the core does not directly relate to radioactive releases as
the high chemical compatibility between the fission products and the sodium coolant trap radionuclides. The
core itself is isolated from the secondary side using an intermediate heat transfer system, which itself limits if
not prevents the propagation of nuclides via being pressurized above the primary sodium loop. A Cover Gas
System is used to remove radionuclides from the cover gas region. Radiation monitors are installed in the head
access area and cover gas service vault to detect any gas leakage. The slightly pressurized argon cover gas is
circulated with a filtering stage to eliminate sodium vapor, aerosols, and any other impurities. This monitoring
system is also used to survey the reactor cover gas to check for elevated fission gas levels that could indicate
fuel failures. The pressurization is used to ensure that no sodium will be released into the environment if a leakoccurs. Argon that is treated is reintroduced to reduce consumption.

Chemical Control
The ARC-100 sodium coolant and argon cover gas are continuously monitored. Both the primary andsecondary sodium coolant are cleaned using sodium cold traps. The primary sodium can also be directed to aradionuclide trap in case of fuel pin failure.

6. Plant Safety and Operational Performances
The ARC-100 will require only minimal active involvement of the plant operators. The operator’s role will be
to monitor plant behaviour and transient response to ensure that it is within the specific design parameters. The
core design features include a low burn-up reactivity swing which reduces the need for frequent control rod
motion and the inherent load following characteristics of the reactor support simplified load following
operation.

7. Instrumentation and Control Systems
The fail-safe safety-related shutdown Distributed Control and Information System (DCIS) is a three-divisioncontrol and monitoring system design, each with separate and independent power supply electrical systems.
Divisions are used to support automatic shutdown of the reactor and decay heat removal via DRACS and225

RVACS. The system is designed to be able to operate with one division continuously out of service when adesign basis event occurs. The DCIS operates at low voltage as the fail-safe shutdown systems are designed tooperate without electricity.

8. Plant Layout Arrangement
Reactor Building
The Reactor Building is a cylindrical building made of reinforced concrete floors and walls. Roof trusses andtheir supporting columns are made of structural steel. The Reactor Building Structure houses the Primary
Reactor System, reactor support and safety systems. The refueling floor of the Reactor Auxiliary Building
Structure includes the refueling and fuel handling systems and the overhead crane. The design pressure andtemperature of the Reactor Building will be established from maximum calculated pressures and temperatures
resulting from postulated design basis events including sodium fires. The reactor building and the primarysystems including the concrete silo will have a design life of at least 60 years.

Balance of Plant
The steam turbine is a non-reheat, air cooled, single shaft, single casing turbine with separate HP/IP/LP
sections. steam turbine with a single flow high pressure turbine and combined intermediate and low-pressure
turbine. The Condensate and Feedwater System collects water from the main turbine and auxiliaries after
available thermal energy in the water has been extracted, conditions it, and returns it to the steam generator atdesign temperature and pressure. The turbine generator auxiliary systems provide supportive services to theturbine generator via cooling, sealing, lubricating, and control functions to sustain the operation and assure themaximum efficiency of the turbine generator.

9. Testing Conducted for Design Verification and Validation
An extensive suite of analysis codes compiled over the 30 years of developmental history of sodium fast
reactors provide a comprehensive theoretical and applied basis for the ARC-100 reactor. Efforts are in progress
to Verify and Validate the codes.

10. Design and Licensing Status
Based on the regulatory requirements applicable to new build projects in Canada, Arc Clean Energy has
successfully completed the CNSC VDR I process. Arc Clean Energy is now completing preliminary design
while progressing through VDR phase II. The site safety assessment, environmental impact assessment, and
PSAR have started.

11. Fuel Cycle Approach
The ARC-100 core uses HALEU in metallic form alloyed with zirconium. The reference plan for the FOAK
demonstration unit is an “open fuel cycle” where spent fuel would be sent to a Deep Geological Repository(DGR). When the reactor is refuelled after 20 years, spent fuel will be moved out of the core region andeventually sent to on-site dry storage. After 60 years of operation and two refuelling cycles, all of the spent
fuel can be transferred to a DGR. The ARC-100 will only produce about 300 spent fuel assemblies over its 60year design life. The ARC-100 reactor is capable of closing the fuel cycle by recycling its own metallic spent
fuel. Given the 20-year fuel cycle and a suitable decay period, recycling the first core load would occur about
50 years after that initial load. The pyroprocessing of spent metallic fuels was developed by ANL anddemonstrated during EBR-II operation to successfully recycle metallic fuel pins. A closed fuel cycle would beconsidered when reprocessing spent fuel is a licensed activity by the regulatory body in Canada, acceptable tothe public, and economically viable. The ARC-100 fuel fabrication process is suitable for shielded productionof recycled fuel.

12. Waste Management and Disposal Plans
The low quantities of solid, liquid, and gaseous radioactive waste resulting from operations will behandled and processed in a responsible and safe manner consistent with the state-of the-art that ensures

minimum exposure to all personnel handling, transporting, and processing the waste. Interim storage will be
stored on the site in defined areas and transported to authorized processing facilities at appropriate times,
dependent on the category and type of waste. Used fuel assemblies are temporarily stored in-vessel at theoutside of the core. Once sufficiently cooled, the used fuel assemblies are periodically extracted and placed
into an on-site dry-cask storage. Long term storage is planned in containers that meet the requirements for the
deep geological repository design of the Nuclear Waste Management Organization (NWMO).

13. Development Milestones
2020

Conceptual Design complete

2023
2026
2027
2029

Preliminary Design complete
License to Prepare Site
License to Construct for first unit
License to Operate First Unit

226

4S (Toshiba Energy Systems & Solutions
Corporation, Japan）
All content included in this section has been provided by, and is reproduced with permission of Toshiba Corporation, Japan.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
Toshiba Energy Systems &country of origin
Solutions Corporation, Japan
Reactor type
Coolant
Thermal/electrical capacity,
MW(t)/MW(e)Liquid metal cooled fast reactor(pool type)Sodium30 / 10

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced convection
Non pressurizedCore inlet/outlet coolant
temperature (oC)355 / 510

Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)

Metal fuel (U-Zr alloy) enriched uranium
18

Reactivity control
mechanismApproach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)

Axially movable reflectors / fixed absorber

Fuel cycle
requirements/approach

Either one-through scheme or closed fuelscheme is applicable
Core lifetime of ~30 years without on-siterefuelling, passive walkaway safety
Detailed design

Distinguishing features
Design status

< 20%
N/A
34

Hybrid passive and active60
157 000
24 / 3.5
Seismic isolator

1. Introduction
The 4S (super-safe, small and simple) is a small sodium-cooled pool-type fast reactor with metal fuel. Being
developed as distributed energy source for multi-purpose applications, the 4S offers two outputs of 30 MW(t)or 10 MW(e) and 135 MW(t) or 50 MW(e), respectively. These energy outputs are selected from the demand
analyses. The 4S is a non-breeder fast reactor. 4S reactor cores are designed to have a lifetime of 30 years forthe 30 MW(t) core and 10 years for the 135 MW(t) core. The plant electric power can be controlled by thewater–steam system, which makes the reactor applicable for a load follow operation mode.

2. Target Application
The 4S is designed for electricity supply to remote areas, mining sites as well as for non-electric applications.
The plant can be configured to deliver hydrogen and oxygen using the process of high temperature electrolysis.
This process can be performed without producing environmentally disadvantageous by-products, such ascarbon dioxide. Two kinds of systems for non-electric applications can be incorporated in the 4S:- Seawater desalination system: the 50 MW(e) 4S plant can produce fresh water at a rate of 168 000 m3/day;- Hydrogen and oxygen production system: hydrogen production rate for the 10 MW(e) and 50 MW(e) 4S is3000 Nm3/h and 15 000 Nm3/h respectively.
Combinations of these systems and the turbine generator system as balance of plant (BOP), including thecapacity of each system, would be determined to meet the actual needs at any particular site.
227

3. Design Philosophy
The 4S reactor is an integral pool type with all the primary components installed inside the reactor vessel (RV).
The 4S design is optimized to achieve improvement of public acceptance and safety, minimization of fuel cost
and O&M cost, use of uranium fuel with enrichment less than 20%, adequate fuel burn-up and reduction incore size.

4. Main Design Features
Power Conversion
The steam turbine generator is used for converting nuclear power to electricity. The nuclear steam supply
system (NSSS) consists of the primary cooling system, the intermediate heat transport system and thewater/steam system. The intermediate heat transport system has an EM pump, piping and a steam generator(SG). The SG is a helical coil type with wire-meshed double-wall tube to prevent a sodium-water reaction in
the event of the tube failure.

Reactor Core
The core and fuel are designed to eliminate the need for refuelling during approximately 30 years for the 10MW(e)-4S and to make all reactivity temperature coefficients negative. Metal fuel, which has an excellent
thermal conductivity, is applied. The core can be operated by axially moving reflectors installed outside of thecore, upward from the bottom. No reloading or shuffling of fuel is required during the whole core lifetime.
The fuel element consists of fuel slugs of U-Zr alloy, bonding sodium, cladding tube, and plugs at both ends.
Coolant inlet modules located beneath the fuel subassembly provide a lower shielding for the reactor internal
structures including the core support plate and air in the reactor vessel auxiliary cooling system (RVACS).

Reactivity Control
The reactivity control during normal operation is by the axial movement of reflectors and using fixed absorbers.
The movable reflector surrounding the core gradually moves, compensating the burnup reactivity loss over the
30 years lifetime. Therefore, the reactivity control is unnecessary at the reactor core side and this is an
important factor to simplify the reactor operation. The transient overpower is prevented by the limitation ofhigh-speed reactivity insertion by adopting the very low speed driving system.

Reactor Pressure Vessel and Internals
The RV houses all the major primary components (integral type) including the IHX, the primary EM pumps,
the moveable reflectors which form a primary reactivity control system, the ultimate shutdown rod which is aback-up shutdown system, radial shielding assemblies, core support plate, coolant inlet modules and fuelsubassemblies. The RV provides a primary boundary for the primary sodium coolant, and is designed with apressure/temperature of 0.3 MPa/550oC. The design lifetime of the RV is 60 years as well as the othercomponents.

Reactor Coolant System
The primary sodium circulates from the EM pumps downward, driven by its pump pressure, and flows throughradial shielding assemblies located in the region between the RV and the cylindrical dividing wall. The coolant
flow changes its direction at the bottom of the RV and then goes upward, mainly into the fuel subassemblies
and partly into the movable reflectors. The coolant flow is distributed appropriately to fuel subassemblies ofeach type and to the movable reflectors. Here, the core barrel separates the core and the reflector regions. Heat
produced in the core is transferred to the coolant while it flows through the fuel pin bundles. The reflectors arealso cooled so that the temperature becomes sufficiently low and the temperature distribution is flattened tomaintain integrity through the plant life time. The coolant gathers at the hot plenum after flowing through thefuel subassemblies and the reflectors. The heated primary sodium then goes into the IHX to transfer heat tothe secondary sodium. During normal operation, the primary system is enclosed inside the RV; sodium coolant
is circulated by two EM pump units arranged in series. The heat generated in the reactor is transferred to thesecondary sodium via the IHX located at the upper region in the RV. The secondary sodium is circulated by
one EM pump unit. The heat is transferred to the water-steam system via heat transfer tubes in the SG. The
heated water/steam is circulated by the feedwater pump.

Secondary System
The secondary sodium loop acts as an intermediate heat transport system and consists of the EM pump, piping,dump tank, and the SG. The secondary sodium coolant heated in the IHX flows inside the piping to the SG
where heat is transferred to water/steam to be supplied to the steam turbine generator.

Steam Generator
The 4S adopts a once through type double-wall tube SG with failure detection systems. The heat transfer tubeof the SG is a double-wall type. Between the inner and outer tube, wire meshes are installed, which are filledwith helium, to detect one side tube failure prior to failure of the other side tube. It enables to prevent sodiumwater reaction.
228

5. Safety Features
Engineered Safety System Approach and Configuration
The philosophy of the 4S safety concepts is to put an emphasis on simplicity achieved using passive andinherent safe features as a major part of the defence in depth (DiD) strategy. In addition to the inherent safety
features, there are two independent systems for reactor shutdown. The primary shutdown system provides fora drop of several sectors of the reflector, and the back-up shutdown system provides for insertion of the ultimate
shutdown rod from a fully out position at the core centre. The reflectors and the shutdown rod are fallen by
gravity on scram. Both the reflector and shutdown rod are each capable of enough negative reactivity to shut
down the reactor.

Decay Heat Removal System / Reactor
Cooling Philosophy
The water/steam system is available for normal
shutdown heat removal. The decay heat of the core istransferred to water/steam system via the intermediate
heat transport system by forced convection and is
finally removed from a condenser. For decay heatremoval during water steam system is not available
upon accidents, two independent passive systems areprovided; the RVACS and the intermediate reactor
auxiliary cooling system (IRACS). The RVACS iscompletely passive and removes decay heat from thesurfaces of the guard vessel (GV) using natural
circulation of air. There is no valve, vane, or damper inthe flow path of the air; therefore, the RVACS is always
in operation, even when the reactor operates at rated
Engineering safety system
power. Two stacks are provided to obtain a sufficientdraft. The IRACS removes decay heat by air cooler which is arranged in series with the secondary sodiumloop. Heat is removed by forced sodium and air circulation at the IRACS when electric power is available. In
addition, the IRACS can also remove the required amount of heat solely through natural circulation of both airand sodium during loss of power events.

Spent Fuel Cooling Safety Approach / System
Spent fuel after 30 years of operation is cooled in the reactor vessel for one year and temporary stored in dry
cask for the 10MWe-4S. Dry cast is cooled by natural convection of air. No need spent fuel pool is required.

Containment System
The 4S adopts a cylindrical/spherical containment system. The containment system consists of the GV and the
top dome, which covers the upper region of the RV, a shielding plug and the equipment located on the shielding
plug. The GV provides the second boundary for the primary sodium at the outer side of the RV. For the
mitigation of sodium fire, nitrogen gas is provided inside the top dome.

Chemical Control
4S does not have chemical control system.

6. Plant Safety and Operational Performances
The 4S is designed to operate safely without active involvement of the plant operators. The design features tosupport such operation include: (1) burn-up reactivity swing automatically compensated by the fine motion
reflectors, (2) no need in reloading and shuffling of fuel in the course of 30 years for the 10 MW(e)-4S, (3)reduction in maintenance requirements achieved by adopting static devices and (4) reduction of in-service
inspections achieved by taking advantage of the non-pressurized systems of sodium-cooled reactor and by
applying a continuous monitoring process based on leak-before-break detection concept to ensure safety.

7. Instrumentation and Control System
The instrument and control system consists of safety related and non-safety related systems. The safety-related
systems include the reactor protection system (RPS), the engineering safety feature actuation system (ESFAS)
and the remote shutdown system (RSS). These systems have the safety class 1E instruments. The RPS is plant
protection system to initiate reactor trip at abnormal plant operation condition.

229

8. Plant Layout Arrangement
The plant layout of the 4S is optimized to meet various
functional needs; the requirements for safety; radiation
zoning, piping and cabling; construction requirements;
and access and security considerations.
(a) Reactor Building
The 4S is a land-based nuclear power station with thereactor building embedded underground for security
considerations and to enhance protection against
extreme external events. The reactor building
including the concrete silo can be used for more than
60 years.
(b) Balance of Plant
The BOP including a steam turbine system is locatedat ground level.
i. Turbine Generator Building
Plant Layout
The 4S plant consists of one reactor and one turbine
generator system. Superheated steam is supplied from the steam generator to the turbine.
ii. Electric Power Systems
These systems include the plant main generator (PMG), the main power transformer and the generator circuit
breaker (GCB), diesel power generator and batteries. The grid is also connected to the unit auxiliarytransformer (UAT). The two class 1E buses are separated from each other and separated from the non-class
1E electric system. Each class 1E system is provided with a separate emergency diesel generator and batteries.

9. Testing Conducted for Design Verification and Validation
Many equipment, fuel and core tests have been performed to demonstrate the separate aspects of the 4S design.
These tests have been performed both in Toshiba’s new sodium test loop facility in Yokohama, as well as
supporting locations worldwide. The following experimental tests for validation were performed:- A critical experiment for the nuclear design method of a reflector-controlled core with metallic fuel.
- The pressure drop for a fuel subassembly has been confirmed by fuel hydraulic testing.
- The scale test of the reflector drive mechanism with fine movement has been performed.
- Heat transfer characteristics between the vessel and the airflow of RVACS have been confirmed.
- The heat transfer characteristics of the steam generator have been confirmed by sodium testing.
- Manufacturability investigations, electromagnetic oscillation testing and sodium flow testing of the EMpump completed.

10. Design and Licensing Status
Licensing activities for the 4S design initiated with the U.S. Nuclear Regulatory Commission (U.S. NRC) in2007. In pre-application review, four meetings had been held in the past and fourteen technical reports havebeen submitted to the U.S. NRC. Toshiba is conducting the detailed design and safety analysis for design
approval. In parallel, Toshiba Energy Systems & Solutions continues to look for customers.

11. Fuel Cycle Approach
The 4S reactor can be applied to either once-through fuel cycle scheme or closed fuel cycle scheme. It mainly
depends on user country’s fuel cycle policy. In the case of once-through fuel cycle scheme, spent fuel after 30
years’ operation is cooled in the reactor vessel for one year and temporary stored in dry cask for the 10MWe4S. Then, it is eventually shipped to a permanent repository.

12. Waste Management and Disposal Plan
Decommissioning at the end-of-life was evaluated such as sodium deposition and reactor vessel deposition canbe done by following decommissioning method of EBR-II and LWR plant in the US. Sodium will be disposedby following experiment of EBR-II. The RV without fuel and sodium will be filled with concrete andtransported to disposal site.

13. Development Milestones
2007
2008
2013

230

Licensing activity for the 4S design initiated with the U.S. NRC
Completion of four times public meetings as pre-application review with the U.S. NRC
Completion of submitting 14 technical reports to the U.S. NRC

MicroURANUS (UNIST, Republic of Korea)
All content included in this section has been provided by, and is reproduced with permission of UNIST, Republic of Korea.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array
Number of fuel assemblies in
the core
Fuel enrichment (%)
Core Discharge Burnup(GWd/ton)
Refuelling Cycle (effective fullpower years)
Reactivity control mechanism

Ulsan National Institute of
Science and Technology(UNIST), Republic of Korea
Lead-bismuth cooled Reactor
Pb-Bi (45-55% Wt.%) eutectic
alloy
60 / 20
Electromagnetic Pump (EMP)
0.1 / 6~12250 / 350
UO2 / Monolithic Inverted
Core85
3 radial zones; 8, 10, 12
60
No refuelling (30)

Approach to safety systems

Control rods and shutdown
rods
Passive

Design life (years)

40

Plant footprint (m2)

N/A

RV height/diameter (m)11.0 / 2.4

RV weight (metric ton)200

Seismic Design (SSE)

3D seismic isolation system

Fuel cycle requirements /ApproachDistinguishing featuresNo HLW/PyroGreen
Recycling
Ex-vessel emergency watercooling system
Conceptual design

Design status

1. Introduction
Based on developments of Lead-Bismuth-Eutectic (LBE)-cooled transmutation reactors such as PEACER,
PASCAR and URANUS, an LBE-cooled micro reactor using UO2 fuel rods has been developed as
MicroURANUS (Universal, Robust, Accident-forgiving, No proliferating and Ultra-lasting Sustainer) with anominal power rating of 20 MW(e) from 60 MW(t). The MicroURANUS is designed to be compact andcompletely sealed after factory-fueled and tested, prior to hot-transportation to installation sites. No refuelingis needed throughout 40-year design life and the encapsulated core will be shipped to recycling facilities sothat no spent nuclear fuels or high-level wastes will be left to the owners. It can operate either in dynamic
environments of merchant ships or in terrestrial settings by utilizing both natural and forced circulation ofchemically inert LBE liquid. For the reliability of system structure components for the design lifetime,necessary design margins are included.
2. Target Application
MicroURANUS will be deployed primarily for the propulsion of merchant ships and for the offshore hydrogenand electricity generation, with zero carbon footprint. MicroURANUS can also be used to replace land-based
231coal plants and to make steep load-follow operations in symbiosis with renewable energy sources.
3. Design Philosophy
Safety in marine environments is assured by eliminating chances of radioactivity release accidents andminimizing the emergency planning zone radius to within engineered power block boundaries. Even in theworst hypothetical accidents, the radioactive core materials will be contained by frozen LBE within the reactor
vessel. Safeguards and physical protection are assured by weld-encapsulated reactor vessel without any
provisions for fuel removal from engineered power block boundaries. It is only after decommissioning whenthe encapsulated reactor module containing spent nuclear fuels can be removed and transported to globally
accepted recycling centers where all high-level waste will be decontaminated to intermediate level wastes forsafe and secure geological disposal without uncertainty. Economy of a MicroURANUS power block will be
far superior to current engines of merchant ships when manufacturing infrastructure is established.
4. Main Design Features(a) Nuclear Steam Supply System
The nuclear steam supply system consists of an integral reactor module that includes a reactor core, twelve
steam generators, control rod clusters, a reactor barrel, a lower plenum, a reactor head, and a reactor vessel.
LBE coolant circulates upward inside the reactor barrel including the reactor core until it meets steam generator
inlet where the flow direction changes to downward by entering stream generator tubes that lead to annular
downcomer formed between the reactor barrel and reactor vessel. The flow direction changes to upward after
passing through a circular flow distributor unit located at the reactor lower plenum.
(b) Reactor Core
MicroURANUS has evolved from URANUS-40 design with hexagonal-lattice using LEU oxide fuels. The
total 85 pellet assemblies, 13 control assemblies and 6 ultimate shutdown assemblies. Three enrichment zones
are used to reduce the radial power peaking and to increase internal breeding gain correspondingly small
reactivity swing. The fissile enrichment adopted for the inner enrichment zone is lower than that of the outer
fuel zone. The MicroURANUS core has an active height of 2.0 m and equivalent diameter of 1.6 m thatgenerates 60 MW(t) with the average coolant temperature rise of 100oC from the core inlet to the core exit.
(c) Reactivity Control
The MicroURANUS has a very small reactivity swing over thirty effective full power years, requiring almost
no control rod movements if operated steadily at full power. Reactivity can be controlled by using 13
assemblies for versatile navigation of merchant ships. Rapid power increases can be tolerated by using PCIresistant fuels. To meet regulatory requirements for the reactor module transportation before and after
commercial operations, additional reactivity suppression is made by using the ultimate shutdown assembly.
(d) Reactor Vessel and Internals
The reactor vessel of MicroURANUS is approximately 11.0 m in height and 2.4 m in diameter. The reactor
vessel is encased in the containment vessel with an annular space used for in-service inspections andmaintenance. The reactor barrel is a cylindrical shell that provides both structural supports for reactor core andinternals while serving as LBE flow divider, neutron reflector and thermal insulator.
(e) Reactor Coolant System
LBE reactor coolant flows up through the reactor core and flows down through steam generator tubes by
hydraulic head generated by both electromagnetic pump (EMP) and natural circulation. The reactor coolant
operates at atmospheric pressure by taking advantage of its high boiling point (1670oC). Dissolved oxygen
concentration in LBE coolant is controlled by injecting oxygen into the upper plenum in order to keep all
structural materials stably passivated against corrosion.
(f) Secondary System
Currently a comparative study is in progress for the secondary systems, between traditional superheated steam
cycle or supercritical CO2 cycle. Criteria for the down-selection process include technical maturity, transient
responses, system footprint as well as maintainability.
(g) Steam Generator
Twelve units of vertical once-through steam generators employ double-walled tubes. LBE flows downward
on tube-side while the secondary working fluid flows upward on shell-side. Inter-tube space is filled withhelium gas for leak detection and thermal conductors. By using ductile austenitic tube materials andcompressive stresses due to higher shell-side pressure, leak-before-break (LBB) of the SG tubes can be
guaranteed and the leakage of the radioactivity in the primary coolant to the secondary side can be prohibited.
5. Safety Features
The MicroURANUS design has been embedded with passive safety features by utilizing high capacity ofnatural circulation and superior heat conduction as well as high boiling point of LBE. In contrast with water,
LBE coolant has excellent retention capability for safety-critical volatile radioactive species including iodine232and cesium. LBE solidification by ex-vessel emergency water cooling can assure long-term isolation ofradioactivity within the reactor coolant system, in the event of hypothetical accidents caused by both internal
events and external events, including flooding, collision, aircraft crashes, explosion, capsizing. Safety featuresof MicroURANUS is designed based on the defense-in-depth principle.
(a) Engineered Safety System Approach and Configuration
Reactor scram systems can be actively actuated with independent emergency power supplies. During normal
shutdown and reactor trip conditions, decay heat is removed by the steam dump through atmospheric dump
valve(ADV) of the secondary steam system when traditional steam system is chosen as the secondary system.
On the other hand, the critical CO2 system is chosen, decay heat is removed by valves the reactor auxiliary
cooling system that circulates coolant to transport heat from the reactor outer wall to external coolers. Under
the beyond design base accident conditions, a dedicated ex-vessel emergency water cooling system is passively
activated to remove any heat beyond the capacity of the reactor auxiliary cooling system. If the reactor vessel
is heated above limits, the annular space between reactor vessel and containment vessel is flooded withshielding water through valves. Steam produced from the water-cooling of the external surface of reactor vesselgoes to dedicated coolers where the steam is condensed into water before returning to the shield water tank by
gravity.
(b) Decay Heat Removal System / Reactor Cooling Philosophy
During normal shutdown and reactor trip conditions, decay heat is removed by the steam dump through theatmospheric dump valve of the secondary steam system when traditional steam system is chosen as thesecondary system. On the other hand, the critical CO2 system is chosen, decay heat is removed by valves thereactor auxiliary cooling system that circulates coolant to transport heat from the reactor outer wall to externalcoolers.
During normal operation and transient conditions, decay heat is removed by the reactor auxiliary coolingsystem that circulates coolant between cooling coils on the reactor exterior to an external heat exchanger.
(c) Emergency Core Cooling System
Under the design base accidents, the passive reactor auxiliary cooling system in combination with heatexchangers submerged in the reactor building pool(shield tank) removes the decay heat from the core, in
addition to this, MicroURANUS has another option that the active AFWS in combination with ADV canremove the decay heat from the core by dumping steam through ADV.
Under severe accident conditions, a dedicated ex-vessel emergency water cooling system is passively activated
to enhance cooling power beyond the capacity of the reactor auxiliary cooling system. If the reactor vessel isheated above limits, the annular space between reactor vessel and containment vessel is flooded with shielding
water through valves. Steam produced from the water-cooling of the external surface of reactor vessel goes to
dedicated coolers where the steam is condensed into water before returning to the shield water tank by gravity.
The ex-vessel water cooling will gradually freeze LBE coolant inward so that radioactive materials can betrapped inside the reactor vessel.
(d) Containment System
The containment vessel and vacuum system forms the containment system. The containment system enclosedby the reactor building pool that is also functioning as the shield water tank. This containment system isphysically protected by external shield from external attacks including aircraft crashes. The containment
vacuum system is replacing the complicated insulation of the conventional nuclear power plants and can also
be used to detect any leakage from the RCS utilizing the radiation detectors in the vacuum system in a shortperiod. The containment system also includes a filtered vent system for venting gas and/or steam when thepressure in the containment increase the pressure limit under the accident conditions.
6. Plant Safety and Operational Performances
The plant safety and operational performance will be assessed by the end of 2022.
7. Instrumentation and Control Systems
The conceptual design of digital hybrid instrumentation and control systems with analogue backup system will
be completed by the end of 2022.
8. Plant Layout Arrangement
The reactor building and the balance of plant will be designed upon the down-selection of the secondary cycle.
9. Design and Licensing Status
A conceptual design of MicroURANUS is to be completed in 2022. Currently, the activities for Front End
Engineering Design (FEED) are carried out for design, optimization, modelling and experimental validations.
Pre-application engagement with the regulatory body will be expected to start in 2025.
10. Fuel Cycle Approach
At the end of life, LEU oxide fuel used for 40-year life will be remain in the reactor for about three years for233

decay heat cooling, prior to decommissioning. The encapsulated core module containing all spent nuclear fuels
will be transported to globally accepted recycling centers, either by land or water. Recycled TRU will be used
to fabricate oxide fuels for the next generation MicroURANUS.
11. Waste Management and Disposal Plan
All high-level waste from recycling of MicroURANUS spent nuclear fuels will be further decontaminatedusing advanced partitioning technology including PyroGreen to intermediate level wastes for safe and secure
geological disposal without long-term uncertainty. Limited quantities of low and intermediate level wastes
produced during operation and from decommissioning can be readily disposed of using available repositories.
12. Development Milestones
2023 ~ 2024
Engineering design and license document preparation
2025 ~ 2027
Pre-application review and finalize the design2028 ~ 2030
Application of the Design Certification and the construction permit of the prototype
reactor for maritime use of MicroURANUS

234

LFR-AS-200(newcleo srl, Italy)
All content included in this section has been provided by, and is reproduced with permission of, newcleo srl, Italy.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
newcleo srl, Italyof originoriginally developed by Hydromine
S.àr.l., Luxembourg
Reactor type
Liquid metal cooled fast reactor (pooltype)
Coolant
Thermal/electrical capacity,
MW(t)/MW(e)Lead480 / 200

Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
0.1 / 18 MPa(abs)Core inlet/outlet coolant
temperature (oC)420 / 530

Fuel type/assembly array
Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)

MOX / hexagonal
61

Reactivity control mechanism

Ex-core, reversed-flag type B4C rods,
rotating B4C Rods

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

Active + passive
60
1 100, nuclear island with fuel storage
6.2 / 662
0.3 g
Stockpiled MOX from reprocessing

Distinguishing features

Active + passive walkaway safety;
No intermediate loops;
Simple, compact primary system: ≤ 1
m3/MWe;Compact reactor building.
Conceptual design

Design status

19% avg / 23,2% max in Pu
16 months (1/5)
100 GWd/ton

1. Introduction
The LFR-AS-200 is an innovative reactor cooled by molten lead; AS stands for Amphora-Shaped, referring to
the shape of the inner vessel and 200 is the electrical power in MW. The embodied innovations exploit thelead properties and enhance the potential for future deployment, owing to plant simplification andcompactness, while behaving passively safe.

2. Target Application
The absence of intermediate loops, the primary system specific volume of less than 1 m3/MWe and the compact
reactor building are key factors for competitive cost per kWh. Market application is energy production withuse of stockpiled Pu and perspective recycle of minor actinides without burden of long-lived transuranics inthe waste. The breeding ratio is 0.9 without blanket assemblies and can be reduced with core design adaptation
where required.

235

3. Design Philosophy
The LFR-AS-200 is a pool-type fast reactor. Main primary components are six innovative spiral-tube steam
generators (STSG), six mechanical pumps, flag-type control rods and three + three dip coolers belonging to
two diverse, redundant decay heat removal systems, fuel assemblies with stem extended above the lead freelevel and hung by their heads. There is no need of in-vessel refueling machine nor of intermediate loops. The
risk of important primary system pressurization, in case of the steam generator tube rupture accident, isdeterministically eliminated by several, special provisions: among them, water and steam collectors locatedoutside the RV and short STSG partially raised above the lead free level of the cold collector.

Primary pumps,
n° 6

Large
rotating
Plug

Small
rotating
Plug

Heat exchangerof the lead-water
DHR, n°3

Lead-lead heatexchangerof the lead-air DHR,
n°3

Spiral-tube
Steam
Generators, n° 6

Reactor Vessel
Safety Vessel

Extended StemFuel Assemblies,n° 61

Flag-typecontrolrods, n°3

Gravityactuated shut
down rods,n°3

Buoyancy
actuated shut
down rods,
n°3

Amphora-Shaped
Inner vessel

Scheme of the LFR-AS-200 and peculiar features

4. Main Design Features
Power Conversion
Superheated water steam is generated in once-through steam generators partially immersed in primary coolant.

Reactor Core
The core consists of 61 wrapped, hexagonal FAs, each containing 390 fuel pins laid out on triangular pitch.
Power shaping or flattening has been achieved through the use of zones with three different levels of Puenrichment. The FAs, the weight of which is supported by buoyancy, present stems extended up to above thefree level, i.e. in gas space. Their heads can be interconnected, and the outer heads fixed also to the section of
ASIV facing the cover gas, by means of cams, which are integral part of each head. The result is a selfsustaining core anchored to the inner profile of the ASIV, that acts as the core barrel in gas space.
The stem of the FAs is peculiar to this novel design, because the FAs heads are directly accessible for handlingwith an ex-vessel refuelling machine operating in gas space under visual control in conjunction with rotatingplugs of classical design. The FA foot is free from mechanical supports (no core grid of classical design suchas the Diagrid of SPX1) except for the radial touching with adjacent FAs, which are brought into contact tocreate a packed bundle, bounded on the lower end by the contoured edge of the bottom port of the ASIV.

Reactivity Control
Core reactivity is controlled in normal operation by ex-core rods, installed in the lead pool between the core
and the ASIV. Buoyancy-driven and gravity driven rods are used as shutdown systems. Ultimate reactor
antireactivity insertion is ensured by core expanders placed on the FAs stems. In case of an ULOOP accident
the strong negative feedback due to radial core expansion – magnified by the expansors placed atop the active
region – quickly reduces the core power to a level that is comparable with the power of the DHR system.
236

Reactor Pressure Vessel and Internals
As an integrated reactor, all the primary components are installed in the Reactor Vessel, among them the key
components: the Amphora-Shaped Inner Vessel (ASIV), the core, the STSG, the Dip Coolers of the Decay
Heat Removal Systems (DHR), and the Recirculation Pumps. The SS316LN stainless-steel is adopted as thereactor vessel material. For reactor internals and fuel cladding, new steels and/or protective coating arenecessary.

Reactor Coolant System
The LFR-AS-200 is a pool-type reactor that uses forced convection by 6 pumps for the lead coolant primarycirculation. The vertical axial-flow pump is integrated inside the SG; the pump rests on, and is connected to,the upper support plate of the SG by means of a flange which closes the pump’s shaft penetration through thereactor roof and supports the variable-speed electric motor of the pump. The pump is characterized by a short,large diameter, tapered hollow shaft containing lead brought in rotation by the shaft itself, in order to increase
the mechanical inertia of the pump. There are no in-lead pump bearings.

Secondary System
The secondary system is based on the Rankine cycle with superheated steam. The only peculiarity is related tothe use of lead as the coolant in the primary system. To avoid lead freezing inside the SG, a prudential highSG inlet temperature is set to 340ºC; this temperature is higher than the melting point of Pb, at 327ºC.
The turbine is made of one high pressure stage and two low pressure stages, with a deaerator fed by steam
from the outlet of the high-pressure stage. The turbo-generator set operates at 3000 rpm. A hot water storageis provided in order to reduce the amount of steam bled from the low-pressure turbine and allow a temporary
operation at 110% Pn.

Steam Generator
The Spiral-Tube Steam Generator (STSG) is an innovative SG conceived for compactness and because it
offers several advantages in term of reactor cost, safety and reactor operability and simplicity of the lead flow
path (Table 1). The SG tube bundle, partially raised above the lead-free level of the cold collector, is composedof a stack of spiral-wound tubes. The inlet and outlet ends of each tube are connected to the feed water header
and steam header, respectively, both arranged above the reactor roof to eliminate, in case of their failure, therisk of large water/steam release inside the reactor vessel. The tube spirals, one spiral for each tube, arearranged one above the other and equally spaced. The SG is fed from the bottom. Hot lead flows radiallythrough the perforated inner shell and, once past the tube spirals, flows into the cold collector through acircumferential window located just below the lead’s free-level.

5. Safety Features
Engineered Safety System Approach and Configuration
Lead has excellent cooling properties and its nuclear properties (i.e., its low tendency to absorb neutrons or toslow them down) enable it to readily sustain the high neutron energies needed in a fast reactor, while offering
the reactor designer great flexibility. Lead has a very high boiling point, namely 1737 °C. As a result, theproblem of boiling coolant is for all practical purposes eliminated. As a coolant operating at atmosphericpressure, the loss of coolant accident (LOCA) can be virtually eliminated by use of an appropriately designed
guard vessel. The LFR-AS-200 exploits lead properties for actuation of passive shutdown and passive decay
heat removal systems, which do not need power sources, operator intervention and logics. Compatibility oflead with air and water allows diversification of the heat sinks: stored water and atmospheric air.

Decay Heat Removal System / Reactor Cooling Philosophy
DHR is performed by means of two diverse, redundant systems, each consisting of three identical loops, eachloop rated 2.5 MW. Two loops are sufficient to remove the decay heat. The loops of the first system are filledwith lead. Each loop consists of a lead-lead dip cooler and of a lead-air cooler with interconnecting piping, andis passively operated and also passively actuated thanks to the thermal expansion of the cold branch of theloop, which actuates the louvers of the air cooler when its temperature exceeds 400°C.

Spent Fuel Cooling Safety Approach / System
The plant is provided a handling flask to transfer the spent fuel from the reactor to the spent fuel water pool. It
is equipped with cooling systems to allow FA cooling during transfer. Cooling is performed in argon forced
circulation. Argon natural circulation will back up forced circulation in case of loss of power.
Spent fuel in “dry” conditions is under consideration.

237

Containment System
The secondary confinement is provided with a concrete containment, external-missile-proof. The dimensionof the containment is kept small because of the low potential energy stored in the coolant (which operates at
atmospheric pressure) and of the small inventory of water/steam of the secondary circuit. Moreover, a safety
vessel eliminates any loss of coolant accident (LOCA), in the event of a failure of the reactor vessel. The“practical elimination” of “core melting” will reduce the design requirements on containment, in particular interm of cooling systems

Chemical Control
An oxygen control system is implemented to avoid the formation of lead oxides and at the same time thepreserve the passivation of steel structures and components. Several systems, already successfully tested on
experimental loops, are under investigation for implementation in LFR-AS-200.

6. Plant Safety and Operational Performances
The safety of the LFR-AS-200 is based on the properties of lead and on the specific design features. One of
the most important characteristics of lead as a coolant is its chemical inertness. Lead is a coolant that does notundergo violent chemical reactions, which could possibly lead to high energy release in the event of accident
conditions. The LFR-AS-200 is designed to operate safely in priority reactor mode i.e. at constant power inthe range 20% and full power. Possibility of reactor load following mode is being investigated, but not
implemented. Load following mode in the range -10% +10% of Pn is possible, by means of adjusting theamount of spilled steam from the low-pressure body of the turbine and using stored hot water as balance heatsink and source. The reactor will operate with an almost stable ΔTcore in the operating range; this imposes theoperation of the MPs at variable speed. Core inlet temperature is maintained constant by control of thefeedwater flow rate. Steam temperature is also maintained constant by the control rods.

7. Instrumentation and Control System
Signals from in-core instrumentation could be made available to both control and surveillance systems.
Moreover in-core instrumentation will remain mostly operational even during refuelling.

8. Plant Layout Arrangement
The reactor building, the spent fuel building and the new fuel building are located on a common basement.
The control room is located above the fuel building. The reactor building extends approximately from 9 m
below grade up to 18 m above grade. The turbine generator building is located at ground level. In addition to
the single-module configuration, two additional arrangements are studied at conceptual level: - a two-modules
configuration with a common turbine generator of 400 MW(e) and - a four-modules configuration with acommon turbine generator of 800 MW(e). A common basement for reactor buildings and fuel buildings isforeseen also in case of multi modules configuration. Economics is expected at turbine generator level but also
by the reduction of the number of spent fuel and new fuel buildings.

9. Testing Conducted for Design Verification and Validation
Testing phase not started, yet. A substantial amount of experimental assessments related to LFR-AS-200
technology will be made for LFR-AS-30 development.

10. Design and Licensing Status
Licensing process has not started, yet. Newcleo is investing in the design of a smaller scale reactor (LFR-AS30) as a FOAK demonstrator, whose licensing is intended to ease the safety assessment of LFR-AS-200.

11. Fuel Cycle Approach
Being a fast reactor, the core of the LFR-AS-200 can be nearly self-sufficient in Pu (breeding ratio ̴ 0,9) or,with some modifications, can also be transformed in Pu burner (breeding ratio ̴ 0,5). A Pu breeder cycle is nota main objective of the project, given the surfeit of Pu available worldwide.

12. Waste Management and Disposal Plan
Uranium, plutonium and minor actinides will be recycled in the LFR-AS-200. Residual fuel waste will betransported to geological repositories.

13. Development Milestones
2014
2019
2021

238

Completion of pre-conceptual design (by Hydromine Nuclear Energy S.àR.L)
Completion of the conceptual design
Hydromine Nuclear Energy S.àR.L. is incorporated in newcleo for a fast development
program. Design restarted after an idle phase

SVBR (JSC AKME Engineering,Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of JSC AKME Engineering, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

JSC Institute for Physics and PowerEngineering and JSC EDB
Gidropress, Russian Federation
Liquid metal cooled fast reactor

Coolant/moderator

Lead-bismuth eutectic alloy

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

280 / 100NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

Low pressure

Number of fuel assemblies in the core61

Forced circulation

340 / 485
UO2 / hexagonal

Fuel enrichment (%)< 19.3

Core Discharge Burnup (GWd/ton)

60 (average)

Refuelling Cycle (years)

7–8

Reactivity control mechanism

Control rod drive mechanism

Approach to safety systems

Passive

Design life (years)

60

Plant footprint (m2)

150 000

RPV height/diameter (m)8.2 / 4.53

RPV weight (metric ton)

280, without core and coolant

Seismic Design (SSE)

0.5g

Fuel Cycle Approach/Requirements

On the first stage will use masteredUO2 fuel with postponed
reprocessing. In the more distantfuture transition to closed fuel cyclewith self-supplying mode
Integral monoblock primary circuit
where reactor, steam generators,
pumps are installed in one vessel
Detailed design for potential
construction in 2025

Distinguishing features

Design status

1. Introduction
The SVBR-100 is a multipurpose small modular fast reactor lead–bismuth (LBE) cooled with an equivalent
electric power of 100 MW. In the Russian Federation, lead–bismuth cooled reactor technology has been usedin several nuclear submarines (NSs). The SVBR technology, according to its basic parameters and salient
technical characteristics, is claimed as a Generation IV nuclear reactor. The development of SVBR-100 isbased on the experience gathered in the design and operation of several LBE facilities on NSs, which allows:- Use of mastered LBE technology;- Use of almost all basic components, units and equipment devices of the reactor installation, which areverified by operational experience in LBE;
- Capability to master primary and secondary circuits;- Use of existing fuel infrastructure;- Ensuring the corrosion resistance of structural materials;- Controlling the LBE quality and the mass transfer processes in the reactor circuit;- Ensuring the radiation safety of personnel carrying out work with equipment contaminated with the 210Po
radionuclide; and- Multiple LBE freezing and unfreezing in the reactor facility.
239

2. Target Application
The possibility of multi-purpose application of modular nuclear power plants (NPP) of different capacities(100 – 600 MW(e)) based on SVBR-100 creates the conditions to satisfy the requirements of consumers in a
new sector of regional and small-scale atomic energy industry: 1) creation of regional NPP and nuclear cogeneration plant (NCGP) of low and medium capacity, 2) utilization as part of floating NPPs, 3) renovation of
NPP units. The standard reactor modules of 100 MW(e) can be used for multipurpose, e.g.:- Modular NPP of small, medium or large power;- Regional nuclear heating and electricity generating plant of 200-600 MW(e) which are located not far fromthe cities;- Refurbish the NPPs with expired reactors’ lifetime; and- Nuclear desalination systems.

3. Design Philosophy
SVBR-100 is designed as a multipurpose modular integral lead-bismuth cooled small power fast reactor to
generate an equivalent electricity of 100 MW(e). The design is based on more than 80 reactor-years operational
experience of LBE cooled reactors for submarine propulsion application. Its main features include:- Enhanced inherent self-protection and passive safety and significant simplification of the design of thereactor as well as entire NPP;
- Possibility to operate with different type of fuel in different fuel cycles (period of operation without
refuelling: not less than 7-8 years);- Compact design and maximum factory readiness of the reactor and its transportability, include railway;- Possibility of creation of module based structured NPP with power multiplying by adding the reactors.

4. Main Design Features
Reactor CoreThe SVBR-100 reactor core operates without any
partial refuelling. The fresh fuel is loaded as a single
cartridge while the spent nuclear fuel is unloaded
cassette by cassette. The core configuration allows
for a lower power density compared with the nuclear
submarines using LBE reactors. This design has thecapability to utilize various fuel cycles. The firststage will be the typical uranium oxide fuel leading
to a core breeding ratio (CBR) of 0.83; MOX fuelcan also be used, leading to a CBR just about 1,
which provides fuel self-sufficient mode in theclosed fuel cycle. Using UO2 as the starting fuel, theclosed fuel cycle can be realized in 15 years. Nitride
uranium and uranium plutonium fuel can also be
used to improve safety and fuel cycle characteristics. The SVBR-100 reactor pursues resistance to nuclear
fissile material proliferation by using uranium with enrichment below 20% while using uranium oxide fuel inthe initial core. The reactor is designed to operate for eight years without core refuelling.

Reactor Coolant System
The entire primary equipment circuit of SVBR-100 is contained within a robust
single reactor vessel. A protective casing surrounds the single-unit reactor vessel.
The reactor transfers generated heat into a two-circuit heat-removal system and SG
with forced multiple-circulation secondary coolant system. Natural circulation of
coolant in the reactor heat removal circuits is sufficient to passively cool down thereactor and prevent superheating of the core. The coolant system includes massexchangers, gas mixture ejectors, sensors of oxygen activity in LBE; its function is
to maintain the LBE quality, inhibiting structural materials corrosion. The circuits
for primary coolant circulation (the main and the auxiliary one), are entirely
realized by components of the in-vessel components, without using pipelines andvalves. Within the main circulation circuit (MCC), the coolant flows according to
the following scheme. Being heated in the core, the coolant flows to the inlet of themedium part of the inter-tube chamber of twelve SG modules connected in parallel
to each other. Then coolant is divided into two flows. One flow moves upwards inthe inter-tube chamber and enters the peripheral buffer chamber with a free surface
level of the “cold” coolant. Another flow moves downwards and enters the outlet
chamber out of which it goes to the channels into in-vessel radiation shielding.
Coolant flows upwards through in-vessel radiation shielding and cools it, and thenit enters the peripheral buffer chamber as well. Out of the peripheral bufferchamber, the coolant flows over the downcomer circular channel along the RMB

240

vessel via the inlet chamber to the MCP suction. Out of the MCP the coolant flows over the two channels
installed in the mono-block of the lower zone of in-vessel radiation shielding into the distributing chamber,from which main part of flow goes to the reactor inlet chamber, thus closing the MCC circuit. Very small partof the coolant moves upwards via the gap near RMB vessel wall, cooling it and goes into the peripheral bufferchamber.

Secondary System
The secondary system includes: SG modules, feedwater and steam pipelines, separators and autonomous
cooling condensers. The basic equipment of SVBR-100 is installed in an 11.5 m high tight box-containment.
In the lower part of each box, there is a concrete well to be flooded by water in beyond the design accidents
that involve failures of all four PHRS via SGs. The reactor monoblock is installed inside concrete well and isfastened on the head ring of roof. In the upper part of the box there is the reactor equipment, including foursteam separators and four cooling condensers immersed in the water tanks PHRS. The high elevation of theseparators has been selected in order to guarantee the coolant natural circulation in the secondary circuit incool down mode. The gas system condensers are installed in the upper part of the box in the separate concrete
compartment.

5. Safety Features
Physical basement for high level of inherent self-protection and passive safety:- First, this is potential energy contained in coolant. At atmospheric pressure, LBE does not store potential
energy, which in an event of accident occurrence can cause destruction of defence barriers, core damage
and disastrous release of radioactivity. For other reactor types and coolants, there are potential energy of
coolant compression and potential chemical energy of coolant’s interaction with structural materials(zirconium) (for water coolant), and with water and air (for sodium coolant);- Potential energy is a natural property of coolant and cannot be changed by any technical solutions;- Further, it is the integral structure of the reactor facility that completely eliminates pipelines and valves withradioactive coolant and eliminates the possibility of coolant leak;- Finally, this is a fast neutron reactor, in which there are no poisoning effects, low burnup reactivity margin,
low value of negative temperature reactivity effect, and negative void reactivity effect. Efficiency of thestrongest absorbing rod does not exceed 0.5$, that being coupled with technical performance of the control
and protection system exclude an opportunity of prompt neutrons criticality in the reactor;- Elimination of radioactivity release into the environment is insured by the system of disposed defence-indepth barriers.
Those type RFs assure their high resistance not only in events of single failures of the equipment and personnel
errors but in events of intentional malicious actions when all special safety systems operating in a standby
mode can be intentionally disabled. At LBE cooled reactors such catastrophic accidents as Chernobyl or
Fukushima disasters as well as fires similar to that occurred at Monju NPP are physically impossible or can beeasily localized with a purpose to prevent population’s exposure to irradiation beyond the NPP site (LOHS
type accidents). This is extremely viable for radiophobia elimination and realization of NPP construction indeveloping countries where the level of terroristic threat is high.
Tank of SG PSHR

Separator
Feed water

Dry saturated
steam

Gas systemcondenser

Pump
Protective device

Reactormonoblock

Reactor vault

BubblerFlooding reactor
vault

241

6. Instrumentation and Control Systems
The main principles of the design include:- Distributed I&C system with several level ofhierarchy and defence in depth;- Soft control of NPP technological systems;- Availability of the large screen and reservezone at main control room (MCR);- Principles of diversity, reliability, physicalseparation and others, providing high level offunctional reliability, including protection
against common cause failures;- Well-developed diagnostic functions; and- Self-diagnostic of I&C programmabledevices.

Design specificities are:- Control of coolant flow rate by changing
rotation speed of MCP depending on reactor
power for maintaining constant coolant heatup;
- Full scope diagnostic system of NPP;
- Providing I&C operability during 7-8 yearsof continuous NPP operation; and- New tasks of neutron flux monitoring while
core refuelling;- Highly reliable reactor control at start up andoperation;- Load-follow operation in the deep range(100–50–100%).

7. Plant Layout Arrangement
General view of pilot plant is shown in Figure.

8. Design and Licensing Status
The Rosatom Scientific and Technical Council convened on 15 June 2006 approved the development of thetechnical design of experimental industrial power unit based on the SVBR-100. Siting licence is received tocurrent time at Dimitrovgrad, in the region of Ulyanovsk. Key reactor and reactor core research anddevelopment works have begun.

9. Fuel Cycle Approach
The reactor without changing the design can operate in various fuel cycles using different types of fuel. On
mastered oxide fuel, core breeding ratio (CBR) will be less than one. On the MOX-fuel, the CBR will beslightly larger than one, and in the closed NFC, the SVBR-100 reactor will operate in the fuel self-supply modewith depleted uranium make-up.

10. Waste Management and Disposal Plan
Spent nuclear fuel (SNF) will be accumulated in the repository until the period when its reprocessing andnuclear fuel cycle (NFC) closure becomes economically viable. Today it is the cheapest fuel cycle.

11. Economic Characteristics
Parameter
FOAK (pilot plant)
NOAK (4 RF )Cost of construction, $/kW(e)~ 6000~ 3000
LCOE, $/MWh
~ 100
~ 60
In accordance with Projected Costs of Generating Electricity, IEA, NEA, OECD, 2015 (discount rate 7%).

12. Development Milestones
2015
2025
2031
2032

242

License for placement
License for constructing (planned)
License for operation and commissioning (planned)
Serial production and supply of packaged equipment (planned)

SEALER-55 (LeadCold, Sweden)
All content included in this section has been provided by, and is reproduced with permission of, LeadCold, Sweden.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of
LeadCold, Sweden
originReactor type
Coolant
Thermal/electrical capacity,
MW(t)/MW(e)Lead-cooled SMR
Lead
140 MWt / 55 MWe

Primary circulation
NSSS operating pressure(primary/secondary), MPa

ForcedNon-pressurisedCore inlet/outlet coolant
temperature (oC)420 / 550

Fuel type/assembly array
Number of fuel assemblies in the core

UN/Hex-can
84

Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup (GWd/ton)

12
300
60

Reactivity control mechanism
Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements / approach

B4C, WB2
Passive28
600 / 20 000 (single unit/fence)
5.5 / 4.8
20
None
TBD

Distinguishing features
Design status

Nuclear battery
Conceptual design

1. Introduction
SEALER-55 is designed by LeadCold for on-grid, commercial power production. It is intended to be fabricatedin series of more than 100 units in an automated factory.

2. Target Application
The primary market for SEALER-55 is existing nuclear power sites, where several units can be clustered toreplace the capacity of existing light water reactors. Other markets to be considered are large consumers ofhigh temperature steam, such as hydrogen and bio-oil/bio-char plants. Remote and marine applications(mining, shipping) may also be considered.

3. Design Philosophy
Since small reactors in general have high specific costs for hardware and personnel, LeadCold has adopted thefollowing provisions to reduce costs for licensing, construction and operation:• Elimination of on-site fuel cycle operations, by designing a long-life core
• Passive removal of residual heat using natural convection of liquid lead and of air.
• Minimisation of reactivity swing, hence control-rod bank worth, by application of 12% enriched UN fuel.

4. Main Design Features
Power Conversion
Rankin cycle.

243Reactor Core
The core of SEALER-55 features 84 fuel assemblies containing 169 UN or (U,Hf)N rods, two B4C control
rod assemblies, ten shut-down rod assemblies using WB2/B4C absorbers and 66 ZrO2 reflector assemblies.
The active height is 1.3 m. The rod diameter, pitch and channel length have been optimised using an analytical
approach published in Annals of Nuclear Energy.

Reactivity Control
Reactivity control is accomplished by two B4C control rod assemblies and ten WB2 shut-down rod assemblies.
The latter are inserted passively by gravity.

Reactor Pressure Vessel and Internals
The primary vessel is manufactured from SS316 with an overlay weld of alumina forming austenite (AFA),
for corrosion protection. The wall thickness is 40 mm, diameter 4.8 m, and height of the cylindrical section
4.3 meters.

Reactor Coolant System
During normal operation conditions, forced circulation of the lead coolant is provided by variable speed reactor
coolant pumps. Eight steam generators transfer the heat to the secondary system.

Secondary System
Rankine steam cycle at 165 bar and T = 350-530°C.

Steam Generator
Spiral tube, where the tube is a compound of Alloy 690 on the steam side and Fe-10Cr-4Al on the primaryside.

5. Safety Features
Engineered Safety System Approach and Configuration
Ten shut-down rod assemblies are parked above the core during nominal operation. These are divided into two
diversified banks, each being capable of taking the core sub-critical by more than 1$. The first consists of highdensity WB2 absorber pellet rods (96% B-10 enriched), which will be inserted by gravity. The second bankconsists of 96% B-10 enriched B4C pellets, which will be inserted by a spring-load mechanism.

Decay Heat Removal System / Reactor Cooling Philosophy
In case the secondary system is unavailable for decay heat removal, dip-coolers will be passively actuated,
permitting removal of decay heat by gravitationally assisted insertion of water at room temperature andambient pressure for a period of at least 200 hours. The dip-coolers are designed to ensure that primary systemtemperatures remain such that restart of the reactor can take place after inspection. In case that dip-coolers failto actuate, a reactor vessel auxiliary cooling system based on natural circulation of air shall ensure that fuelrod cladding temperatures never exceed their rapid creep failure limit. However, primary vessel creep may
occur, and restart of the reactors is not foreseen.

Spent Fuel Cooling Safety Approach / System
The fuel resides in the primary system during the entire life of the plant. Following shut-down of the plant, thefuel will cool in-situ until the primary system is removed from site.

Containment System
The SEALER reactor unit is located underground with a concrete top plug for airplane crash protection, andas such does not require a conventional containment in form of a biological concrete shield. The steel
confinement of the nuclear island is designed for an overpressure of 0.4 MPa.

Chemical Control
An oxygen control system consisting of oxygen sensors and oxygen pumps is ensuring that the concentrationof oxygen in the primary coolant is such that lead oxide precipitation will not occur, and that a sufficientamount of oxygen is available for self-healing of any damage that may be inflicted on alumina forming steel
surfaces. All surfaces exposed to lead consist of alumina forming steels (ferritic, austenitic or martensitic)
which have been proven to be highly corrosion tolerant.

244

6. Plant Safety and Operational Performances
SEALER-55 is designed to provide a passively safe, secure and reliable power source on-grid or industrial site
applications. The reactor is able to produce 55 MW of electric power for 25 full power years without reloading
nor reshuffling of its UN fuel. A capacity factor of 95% is foreseen, permitting preventive maintenanceexchange of pump and steam generator modules with a frequency of 1/year. No fuel reload is foreseen for theentire life of the reactor (25 full power years). The application of alumina alloyed steels provides corrosion
protection that is deemed sufficient over the life of the reactor. Safety analysis shows that as designed,
SEALER can survive unprotected loss of flow, loss of heat sink and transient overpower accidents with noconsequences for fuel and clad integrity. Moreover, the source term is sufficiently small that a full release ofvolatile fission products into the coolant at the End of Life (EOL) does not require permanent relocation fromhousing residing beyond 1.0 km from the point of release.

7. Instrumentation and Control System
The primary system is instrumented in order to measure neutron flux, temperature, oxygen concentration andhot leg lead free level position.

8. Plant Layout Arrangement
The Reactor Building is located below grade and contains the primary system (reactor vessel) and all primary
auxiliary systems that can contain radioactive material. The systems are located in the confinement structure,
which consists of a steel wall enclosing the reactor hall, connected to the guard vessel. In addition, areas areprovided for storage of used activated components such as steam generators and pumps. The confinement
system is designed for 0.4 MPa overpressure. The concrete building structure does not serve a confinement
function for radioactive materials. It contains a top plug designed to provide protection against external
hazards, such as aircraft impact.

9. Testing Conducted for Design Verification and Validation
An electrically heated prototype of SEALER-55 will be built and operated in Oskarshamn, Sweden for the
purpose of validating its safety concept, operational and maintenance procedures, as well as materials
performance. The prototype is designed with a power of 3 MW, produced in 7 heated rod assemblies with 37
rods each. The height of the prototype is 1:1 with respect to the SEALER-55, in order to validate residual heatremoval capability by dip-coolers, whereas the diameter of the vessel is scaled by 1:2.5. The prototype iscurrently under engineering design, with the intent to have it in operation by 2024.

10. Design and Licensing Status
Interactions with the Swedish Radiation Safety Authority (SSM) have taken place, and a regulatory path
towards licensing of a lead-cooled research/demonstration reactor in Sweden has been identified. This path isbased on a combination of IAEA guidelines for research reactors, and existing regulation for light water
reactors, when the latter might be applicable. Moreover, the Swedish government has given the task to SSM
to develop a framework for licensing of SMRs.

11. Fuel Cycle Approach
The uranium nitride fuel of SEALER-55 will be fabricated by direct ammonolysis of 12% enriched UF6,
followed by spark plasma sintering of pellets.
The spent fuel of SEALER-55 reactors may either by disposed in a geological repository, or recycled. In theformer case, disposal of entire cores in frozen lead, conversion of irradiated UN to UO2 pellets or direct
disposal of UN fuel rod assemblies will be evaluated.

12. Waste Management and Disposal Plan
See fuel cycle approach.

13. Development Milestones
2018

Concept design of SEALER-UK funded by UK BEIS

2020

SUNRISE project, including design of a demonstration unit funded by SSF.

2021

LeadCold and Uniper form joint venture SMR AB

2022

The SOLSTICE project including design, construction and operation of the electricallyheated mock-up SEALER-E in Oskarshamn is funded by the Swedish Energy Agency.

2024

The SEALER-E electrical mockup is taken into operation

2030

The SEALER-D nuclear demonstration unit is taken into operation

245

246

Westinghouse Lead Fast Reactor(Westinghouse Electric Company, USA)All content included in this section has been provided by and is reproduced with permission of Westinghouse Electric Company LLC, USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type
Coolant/moderator

Westinghouse Electric Company,
LLC, U.S.A.
Pool-type, liquid metal cooled
fast reactor
Lead / fast spectrum

Thermal/electrical capacity,
MW(t)/MW(e)Load following (MWe)Primary circulation

950 / ~450 (Net Avg.),
300 - 600 with constant core
power
Forced circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

Nearly atmospheric / 33

Number of fuel assemblies in
the core
Fuel enrichment (%)
Refueling Interval (full power
years)
Core discharge burnup(GWd/ton)
Reactivity control mechanisms

390 / 650 (target, pending
material qualification)
Oxide, with provision fortransition to Nitride325≤ 19.75%
8 - 15 (depending on fuel)up to 100

Design life (years)

Control and shutdown rods;
thermal/pressure activated, nonrod-based passive shutdown
Passive: IAEA passive safety
category B goal
60 (components), 100(structures)

Plant footprint (m2)

~40 000 w/ air-cooled condensers

RPV height/diameter (m)Approx. 9.0 / 6.7

RPV weight (metric ton)

~215 (Assembly of RV, GV,
flange, and passive coolingpassages)
The same as typical LWRs

Approach to safety systems

Seismic Design (SSE)
Fuel cycle
requirements/approachDistinguishing features

Design status

Once-through fuel cycle withsingle-batch long-life core
design. Closed cycle possible.
Compact configuration withhybrid microchannel heatexchangers; non-reactor-basedload follow
Conceptual design

1. Introduction
The Westinghouse lead fast reactor (LFR) is a medium-output, modular construction plant harnessing a leadcooled, fast spectrum core operating at high temperatures in a pool configuration reactor. The simplicity of its
safety systems, high-efficiency advanced ultra-supercritical (AUSC) water balance of plant (BOP), compact
reactor building, absence of high-pressure containment, streamlined modular construction, and integrated nonreactor-based load following capability results in unparalleled economic potential, allowing the Westinghouse
LFR to supply clean competitive energy even in the most challenging market conditions.

247

2. Target Application
The Westinghouse LFR is designed to be a versatile plant, with baseload electricity production and load
leveling as the primary design focus, but with the capability to fulfill a range of non-electric applications suchas process heat, desalination, and hydrogen production needs according to market demand. Its output issufficiently small to integrate lower-capacity grids while also being substantial enough to be used in standard
baseload plant applications. High temperatures permit extremely high BOP efficiencies and use for many
process heat applications. Furthermore, a lower cost per MW(e) permits further electrical boosting of processtemperatures while still being competitive at any temperature. Similarly, high temperature and BOP efficiency,
combined with projected low plant cost, permit the use of hybrid heat/electricity methods providing thepotential for cost-effective hydrogen generation. Integrated thermal energy storage using low-cost materials,
coupled to standard BOP equipment, allows for non-reactor-based load follow to complement nondispatchable energy forms while maximizing energy production. These capabilities could allow for theincreased use of renewable technologies, making nuclear power and renewables complementary.

3. Design Philosophy
The Westinghouse LFR was designed to harness the outstanding safety, neutronic, and thermal characteristics
of molten lead coolant to simplify safety, reduce overall plant size, and maximize the BOP efficiency. This,
when coupled to fuel cycle flexibility resulting from operation in fast neutron spectrum, and non-reactor-basedload follow, drives towards the ultimate goals of being economically competitive against any competing energy
form in free markets while maintaining mission flexibility to customers worldwide.

4. Main Design Featuresa) Balance of Plant (Power Conversion)
While previous variants of this technology aimed to take advantage of an advanced supercritical CO2 power
cycle, the unique operating conditions and needs of the LFR, along with advances in steam cycles (so-calledadvanced ultra-supercritical technology), has resulted in a technology change “back” to a water-based cycle.
While this does result in larger turbomachinery, it permits shrinking of the primary heat exchangers and theirconnections on the LFR’s highly compacted vessel layout. Along with other LFR-specific benefits and reduced
capital cost, it has allowed a significant reduction in the LFR’s vessel size and unprecedented power density.
Furthermore, due to the aforementioned cycle developments, over 47% net efficiency is predicted using an aircooled condenser (ACC).

b) Reactor Core
The core employs a conventional fuel assembly configuration, featuring solid fuel in cylindrical cladding,arranged in a hexagonal array. A staged approach to deployment is adopted for fuel system materials, featuring
higher-maturity materials for LFR technology demonstration followed by higher performance materials as they
become available for use. Specifically, higher technology readiness oxide fuels (UO2 and MOX) in 15-15Titype austenitic steel cladding, such as D9, are considered for near-term use, while advanced fuel (uranium
nitride being the focus) in an enhanced-performance cladding are envisaged to support an increase in burnup
and operational temperature in follow-on LFR units. Consistent with LFR’s goal to ensure fuel and fuel cycle
flexibility, fuel transitions can easily be accommodated as the various core designs only differ in geometriccharacteristics inside of the assembly duct, with assembly footprint and core layout being the same regardlessof the fuel adopted. The core design uses 235U fuel with enrichment <20% while maintaining the option to burnPu-containing fuel through its MOX version.

c) Reactor Coolant System
The Westinghouse LFR features a novel reactor design configuration utilizing high power density hybrid
microchannel-type primary heat exchangers (PHE) integral to the upper part of the core barrel. The
compactness of the PHE design reduces the overall volume of the RV. This arrangement allows the reactor
coolant pump (RCP) impellers to be placed in the lower temperature coolant discharging from the PHEs,
reducing service temperature of the rotating components and related material design challenges. The figure
above depicts the primary coolant flow path. After exiting from the PHE, primary coolant at cold pooltemperature flows through the RCPs and is sent to the core, where it is heated and discharged to the upper
plenum. The primary coolant then flows radially through the PHE and returns to the RCP inlet plenum. The
configuration and resulting flow pattern also ensure the entire RV is in contact with lower temperature coolant,
easing material requirements for this component.

d) Primary Heat Exchangers
Located inside the reactor vessel and integrated into the core barrel internal structure are the three primary heatexchangers (PHEs) to transfer heat to the secondary side working fluid; supercritical H2O at 330 bars. With nowelds in the main body, very small H2O channels within diffusion bonded plates, and fluid headers locatedoutside of the RV/contained area, a robust structure capable of maintaining extreme pressure differentials iscreated. When combined with the lack of exothermic reaction between primary lead coolant and BOP fluid,these elements allow PHE’s placement directly into the RV pool with limited risk of a significant RV
248

pressurization event, eliminating the need for an intermediate heat transport loop present in other advanced
reactor technologies, resulting in a more compact and cost competitive plant.

e) Non-Reactor-based Load Levelling
Westinghouse is currently developing thermal energy storage systems capable of providing load-levelling forthermal power plants. The storage system maximizes economic advantage by being integrated with the same
turbine and generator as would be used for power generation, managing supply fluctuations produced byrenewable sources by storing heat energy when electricity demand is low and selling produced electricity fromthat stored heat when it is high, all while maintaining the reactor core at full power. This is accomplishedthrough manipulation of feedwater and turbine extraction flows to either increase or decrease the mass flow
and steam conditions through the turbine.
At present, it is predicted to allow non-reactor-based load follow continuously between 300 and 600 MWe(Net) with round trip efficiencies competitive with battery solutions. In addition to ongoing work to enable
coupled electricity + heat, hydrogen generation, and integrated desalination, it should be noted that the same
systems and components used in the energy storage system may be integrated with solar-thermal boosting..

5. Safety Features
The Westinghouse LFR harnesses the inherent favorable safety characteristics of the lead coolant to simplify
the reactor design and lower plant cost while allowing for the highest level of safety.

a) Robust, Inherently Safe Design Characteristics
The following characteristics of the design enhance its safety inherently:
 Thermophysical properties of lead, including its high boiling point (1745°C); atmospheric pressure
operation; lack of exothermic chemical reaction with water and air; ability to retain some key fission
products; shielding capability; high thermal conductivity; and, when combined with a pool-type primarysystem configuration, high thermal inertia
 Lead’s excellent neutronic properties for operation in fast neutron spectrum allows the fuel rod lattice tobe opened relative to sodium fast reactors, resulting in a minimum neutronic penalty while providing asignificant enhancement in natural circulation capability during accidents, due to the associatedreduction in core pressure drop Integral, pool-type configuration of the primary system eliminates primary line break, thus eliminating
loss of coolant concerns by design
 Favorable reactivity feedback typical of liquid metal fast reactors
 Robust, microchannel hybrid PHE reduces chance of secondary break and substantially reduces its
severity Underground placement of components important to safety.

b) Passive Heat Removal
The LFR harnesses its high temperature capability in order to use radiation heat transfer between the RV and
GV to remove reactor core decay heat during a safety event. The GV is submerged in a pool of water sufficient
to remove heat from the reactor for several days, subsequently transitioning to natural circulation air-coolingonce water is depleted. Being radiation heat transfer-driven, this system does not require any actuation signal(i.e., it is always on) and does not result in significant parasitic losses during normal plant operation as thecorresponding RV temperature, on the order of 400°C, is too low to result in effective radiation heat transfer.
Instead, during a safety event the increase in RV temperature accelerates radiation heat transfer which, becauseof its function of temperature to the 4th power, becomes effective in removing RCS heat only when it is needed.

c) Pressurization Events
In addition to not having any source of credible RV pressurization events originating from the BOP, no BOP
plenums, piping, or headers will be located within the primary containment (represented by the RV, GV andcommon cover), thus removing the potential for large leaks in nuclear-related areas. The limiting credible
break size in the contained area is reduced to the microchannels in the diffusion bonded block of the PHE.
While small in terms of leak size, the RV pressure resulting from a microchannel break could become
substantial over time without mitigation. Due to the target of IAEA Passive Safety Category B, no isolation
valves are credited for use in PHE leaks (although non-safety isolation valves will be present). Instead,condensing chambers are paired with a passive, pressure-activated reactor shutdown mechanism and tailoredpower conversion system design which limit releasable inventory, ensure RV integrity, and eliminate the needfor a traditional, high-pressure containment.
Reactivity management and shutdown. A reliable, diverse, redundant reactivity control and shutdown systemcomprised of control and shutdown assemblies ensures reactivity management and shutdown capabilities forall operating and anticipated events. This includes the possibility for a non-rod based, passive means to shut
down the core based both on over-temperature and over-pressure transients.

249d) Spent Fuel Approach
Because of, primarily, adoption of a longer active length relative to traditional fast reactors, the LFR core hasa low power density (despite the high-power density of the overall nuclear system). Among many otherbenefits, this low power density has commensurate reductions in linear decay heat power. Recognizing thelong, unshuffled, once-through fuel cycles possible with the LFR and their intrinsic capacity factor benefit,
this low duty condition supports adoption of a direct-to-cask refueling scheme, with continuing work being
conducted to prove its safety and commercial efficacy.
Utilizing Westinghouse’s latest cask technology, this system will allow for direct lifting of used fuel assemblies
into a cask following a brief waiting (cooling-off) period, estimated at less than 90 days. Given the infrequencyof refueling, especially in anticipation of further advanced fuel types (such as Uranium Nitride), the many
benefits in plant and operational simplicity, and reduced capital cost, were deemed worthwhile to pursue inexchange for extended, although infrequent, outages.

6. Instrumentation and Control Systems
A design goal for the development of the Westinghouse LFR’s plant safety systems is to not rely on signals
from the instrumentation and control (I&C) system. As a result, most components, systems, and software used
to control the plant will be commercial grade.

7. Plant Layout Arrangement
As previously noted, the Westinghouse LFR design results in an ultra-compact nuclear system such that mass(often a challenge for lead-cooled reactors) is not such as to require non-standard support provisions. The GV
and RV will be suspended from concrete support pads into a safety pool. The safety pool into which the GV
is submerged will reside in the lowest levels of the plant, with no adjacent rooms or cavities to form leakage
paths. All of these areas, as well as an area located above the reactor platform, will be located underground.
At grade elevation will be an impact shield sufficient to provide protection from external threats. Allcomponents and systems with safety significance (outside of the safety pool itself) will be part of the fabricated
RV/GV assembly and attached condensing tanks (located in safety pool), thus truly providing a ‘nuclear
island.’The use of an AUSC system on the LFR reduces the size of in-vessel primary heat exchangers and theirconnected lines significantly and eliminates the need for a large condenser to be located in the turbine building(below the turbine). This shrinks the RV and turbine building and greatly eases the structural requirements ofthe turbine building, as the turbine can be placed much closer to grade. This arrangement results in a compact
nuclear/turbine island with significant, large-component erection (of the ACCs) performed outside of the plant,
allowing for more parallel construction activities and reduced construction duration.

8. Design and Licensing Status (Design and Testing Status)
Westinghouse has established an international team of partners dedicated to successfully delivering a highperformance commercially viable plant. Between ten and twenty state-of-the art test facilities in the U.S., U.K.
and Europe are currently being used or constructed in order to demonstrate key LFR’s materials, systems,
components and phenomena. These include a full-height passive heat removal facility, a versatile lead loop
facility to test components such as the fuel bundle and PHE at prototypic conditions, and multiple material
testing facilities.

9. Development Milestones
Relatively conventional materials have already been extensively tested in liquid lead and/or in fast spectrum,e.g., SS316 for low-DPA structural components and D9-type steel for fuel rod cladding. Testing of moreadvanced materials, and subsequently of individual components at higher temperatures, will (and is being)
performed in a controlled environment to qualify them for use in evolved designs. LFR serves to answer thecall of future energy markets, allowing a multitude of missions, fuel cycles, and operating strategies to be
adopted in various locations and nations around the world.

10. Development Milestones
2020s – 2025

2025 – 2026
2030
Beyond 2030s

250

Design development and licencing process
Additional testing is performed to validate enhancements to the plant performance.
Start of construction of the first of a kind (FOAK) LFR demonstration plant
FOAK LFR demonstration plant to generate ≤300 MW(e)A higher output ~450 MW(e) FOAK plant representative of the commercial fleet will belicensed and deployed.

PART IV. MOLTEN SALTSMALL MODULAR REACTORS

251

252

IMSR400 (Terrestrial Energy Inc.,
Canada)All content included in this section has been provided by, and is reproduced with permission of, Terrestrial Energy Inc., Canada.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
Terrestrial Energy Inc., Canadaof origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa

Molten salt reactor
Fluoride Fuel Salt/graphite
440 / 195 per operating Core-unit; twoCore-unit configuration
Forced<0.4 (hydrostatic)Core Inlet/Outlet Coolant
Temperature (oC)

620 / 700

Fuel type/assembly array
Power conversion process

Molten salt fuelSteam generator supplying superheated steam to turbine<5%84; before Core-unit replacement
Short-term: negative temperature
coefficientLong-term: online fuel addition
Passive

Fuel enrichment (%)
Refuelling cycle (months)
Reactivity control mechanism

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements /approachDistinguishing features

Design status56
45 00018.0/4.1300
0.3g
Once-through, low enriched uranium
fluoride eutectic mix Fuel Salt
Core-unit replaced completely as asingle unit every 7 years; Fuel salt isreused from one Core-unit to the next
Detailed design

1. Introduction
The integral molten salt reactor (IMSR®) is a 442 megawatt-thermal per Core-unit (IMSR400) small modular
molten salt fuelled, graphite moderated, thermal spectrum reactor. Terrestrial Energy has developed a two-unit
configuration which can deliver 884 MW(t)/390 MW(e). The IMSR® integral nuclear reactor design. features
a completely sealed reactor vessel with integrated pumps, heat exchangers and shutdown rods all mounted
inside a single vessel; the IMSR® Core-unit. The sealed Core-unit is replaced completely at the end of its
useful service life (nominally 7 years). This allows factory production levels of quality control and economy,
while avoiding any need to open and service the reactor vessel at the power plant site. The IMSR400 achieves
the highest levels of inherent safety as there is no dependence on operator intervention, powered mechanicalcomponents, coolant injection or their support systems such as electricity supply or instrument air in dealing
with upset conditions.

2. Target Application
The IMSR® plant is designed to serve electric power grids and industrial heat and power application as
represented in the schematic.

253

3. Design Philosophy
The underlying molten salt
technology of the IMSR®
design is molten salt
reactor technology thatwas the product of
successful and extensive
research programs at the
Oak
Ridge
National
Laboratory (ORNL) in the
1950’s, 1960’s and 1970’s
Terrestrial
Energy’s
technology development
focuses on building thefoundations of cost and safety with ORNL’s design as a starting point. Three technical programs are tightly
integrated – R&D, Engineering, Safety Analysis – over the entire project life-cycle design, construction,commissioning, operations, and decommissioning. These programs and the attendant processes and outputs(documentation in accordance with Terrestrial Energy’s Management System) are the feedstock for licensing
submissions and to the Cost Engineering program as well).

4. Main Design Features
Power Conversion
The solar salt loop, which is pumped from the nuclear island to a separate building outside the Nuclear Facility,
either supplies a steam generator that generates superheated steam for power generation or is used to drive
process heat applications. The steam circuit powers a conventional, off-the-shelf industrial steam turbine forpower generation and/or industrial steam production, depending on the required application. Alternately, someor all the hot molten solar salt may connect directly to a process heat application.

Reactor Core
The Core-unit is manufactured in a controlled factory environment and then brought to the reactor plant sitewhere, following final assembly, it is lowered into a surrounding guard vessel located in a below grade reactor
silo. The IMSR® Fuel Salt is a thermally stable fluid with excellent coolant heat transfer properties, and highintrinsic radionuclide retention properties. As shown in the above figure, a secondary coolant salt loop, filled
also with a fluoride salt (but without fuel), transfers heat from the Core-unit Fuel Salt via primary heatexchangers to a third industrial solar salt loop.

Reactivity Control
Reactor control is assured through negative temperature feedback made possible by the neutronic behaviourof the molten salt fuel in the reactor core. This negative temperature feedback avoids overheating by assuring
criticality control, without the need for the activation of the engineered control system. Molten salt fuel does
not degrade by heat or radiation, which gives a high-power limit to the salt fuel. Although shutdown rods areintegrated into the IMSR® Core-unit, these are for the operational control, and not needed for safety. These
shutdown rods will shut down the reactor upon loss of forced circulation and will also insert upon loss of
power.

Reactor Vessel and Internals
The Reactor Vessel houses the graphite core, the integral primary pumps and heat exchangers, and the thimbles
that receive the shutdown mechanism rods.

Fuel Characteristics
The IMSR® is a liquid-fuel reactor – there are no solid fuel elements in the reactor core. The fuel, in the formof uranium tetrafluoride (UF4), is infused with a primary coolant forming a near-eutectic salt mixture. The
primary salt mixture consists of low-cost fluoride salts and avoids the use of either lithium or beryllium. The
benefit of this primary coolant is that it minimizes the production of tritium. The fuel- primary coolant salt mix
is pumped through a critical, graphite moderated (thermal spectrum) core, and then through the integral heatexchangers to transfer its heat to the secondary coolant salt loop. Overall, the IMSR400 Fuel Salt provides thebasis for a less complex reactor configuration with many safety attributes.

Fission Product Management
The first barrier for fission product release is the fluoride Fuel Salt. Under the operation redox potential, themajority of elements produced as a result of fission (fission products) will form their respective stable fluorides,
trifluorides, and iodides, which are soluble in the Fuel Salt. The Core-unit/Reactor Vessel represents the second
barrier. Since the Core-unit operates at low pressure the probability of explosive release is extremely low.
Additional successive barriers are the Guard Vessel, the Containment, and the Reactor Auxiliary Building
254

(RAB).

5. Safety Features
Engineered Safety System Approach and Configuration
Safety management of IMSR® technology is less complex than Conventional Nuclear reactors and is foundedon the three concepts of nuclear safety previously mentioned – Control the reactor, Cool the fuel, and Containthe radioactivity. The IMSR® design achieves these goals by taking advantage of the passive and inherent
attributes of IMSR® technology that are set out in the figure below.
Control

Cool

Contain

IMSR® has a strong negative IMSR® design assures heat IMSR® Fuel Salt forms strong ionicreactivity
coefficientof dissipation in all circumstances as chemical bonds with many fission
temperatureforinherent IMSR® Fuel Salt is molten and products to uniquely containreactivity control.
serves also as the coolant. This radioactivity within the salt and theuniquelyenablesconvective reactor.
Passive shutdown that does not cooling of the nuclear fuel.
require the traditional use of
No chemical driving forces arecontrol rods.
IMSR® design is small and present, in part, as zirconium metaloperates at 700oC. Small high water reactions are not possible.
temperature reactors have greater
No physical driving forces as IMSR®
inherent cooling capabilities.
operates at atmospheric pressure.

Decay Heat Removal System
The decay heat from the fuel in the reactor core is removed by the Internal Reactor Vessel Auxiliary Cooling
System (IRVACS). IRVACS is an air coolant based fully passive system with no operator intervention required
and does not rely on any instrumentation for its operation.

Emergency Core Cooling System
The IMSR400 is designed with passive heat removal design features; therefore, an Emergency Core Cooling
System is not required for safety.

Containment System
The Containment system consists of a leak-tight fully metallic envelope that houses radioactive materials. This
includes radioactive materials (irradiated Fuel Salt and Off-gases) in the Core-unit, Fuel Salt Storage Tanks(FSSTs), Gas Holding Tanks (GHTs), Hot Cell, and connecting piping.

Chemical Control
Chemical control is achieved by ensuring certain predetermined purity of materials (fuel-bearing fluoride salt
mixture, graphite and alloy) prior to operations and establishing and maintaining the predetermined redox
potential window inside the Core-unit as well as the prevention of material contamination during operations.

Spent Fuel Cooling Safety Approach / System
The irradiated Fuel Salt will reside in Fuel Salt Storage Tanks which are cooled by a combination of passive
and active design features. After the decay heat reaches a sufficiently low temperature, the Fuel Salt istransferred to spent fuel storage tanks which are cooled passively through heat transfer to the ambient.

6. Plant Safety and Operational Performances
The IMSR technology meets the safety principles established by the Global International Forum for Generation
IV nuclear energy – “operations will excel in safety and reliability.” The quantitative safety goals are alignedwith the Canadian Regulator’s (the CNSC) expectations. Due to its molten Fuel Salt characteristics, theIMSR400 “core damage frequency” safety goal of less than 10-5 per reactor year is expressed in terms of lossof reactor vessel integrity. The large release frequency is less than 10-6 per reactor year. The plant design
incorporates automation features to maximum extent possible and the relative simplicity reduces need for large
staff complements. The plant layout is optimized to accommodate nuclear safety features such as fire
protection, radiation protection, security, and safeguards. The IMSR400 capacity factor is 95% with 128 outage
days planned for Core-unit replacement (every 7 years) and other maintenance.

7. Instrumentation and Control System
The control safety function for reactor power for the IMSR400 is inherently part of the reactor’s fundamental
physics characteristics and does not require an engineered active system to maintain control of power in any
design basis accident scenario. A plant investment protection system and a plant control system provide all255control functions.

8. Plant Layout Arrangement
The IMSR® Plant consists of two
distinct facilities: a Nuclear Facility(NF) and a Thermal and Electric
Facility (TEF) shown in the figure
below. Acting together, they form the
IMSR® Plant, an 884 MW(t) nuclear
powered cogeneration facility. The
TEF output can be configured in any
combination of thermal and electric
power output. In the figure, the NF
shows two Reactor Auxiliary
Buildings (RABs), R1 and R2, eachhousing one operating Core-unit within its own Containment system. A common Control Building is locatedbetween the two RABs.

9. Testing Conducted for Design Verification and Validation
TEI has completed the first phase of testing to confirm Fuel Salt thermo-physical properties. The tests
accounted for Fuel Salt aging - build-up of fission products. Verification of the experimental findings’
reproducibility will follow. Testing of Fuel Slat redox potential and respective interfacing material response
is in-progress. Solubility limits of trifluorides and iodides inside the salt are in progress. Salt chemical andmechanical interactions with non-irradiated graphite is completed. Analogous test using the irradiated graphite
specimens will follow. Irradiation of selected graphite grades is in progress and will conclude with theproperty-testing of the irradiated graphite. The test devoted to the irradiated alloy property-testing entered its
detailed design phase. Alloy corrosion testing is in progress. Waste management feasibility study is completed,and experimental phase is being planned. Reactor physics, Thermal-hydraulics and instrumentation tests arebeing designed and planned.

10. Design and Licensing Status
Terrestrial Energy completed Phase 1 of the Canadian Nuclear Safety Commission’s (CNSC) Vendor Design
Review (VDR) in November 2017, successfully meeting CNSC’s requirements for the IMSR® design. In
October 2018, the IMSR® entered Phase 2 pre-licensing VDR with the CNSC. Phase 2 VDR involves adetailed follow-up on Phase 1 VDR activities, and an assessment of the IMSR® power plant design’s ability
to meet CNSC’s expectation in all its 19 design focus areas. VDR Phase 2 submissions have been completed
as of early 2022 and the CNSC is preparing the final report for issuance. Regulatory interface with US NRC
commenced in 2019 and is progressing toward first topical reports in late 2022. Terrestrial Energy’s Postulated
Initiating Events (PIE) methodology was subject of joint review by CNSC and US NRC, and the Regulators’
final report was published in 2022.

11. Fuel Cycle Approach
The IMSR400 fuel consists of low enriched uranium fluoride, in a eutectic mix with other fluoride salts andutilizes a once-through, low enriched uranium fuel cycle. During Core-unit’s operation, makeup Fuel Salt is
added online. The development of the fuel supply chain is in progress with key suppliers for uranium
conversion and enrichment services, and for transport and packaging. At the end of its useful life, the spent
fuel is stored onsite until it can be moved offsite to a final waste management facility.

12. Waste Management and Disposal Plan
The used fuel from each Core-unit at its end of life is recycled/used in the next Core-unit through the plant’slife cycle. After the end of the last Core-unit’s life, the spent fuel is stored onsite until it is ready fortransportation to and storage in an offsite final high-level waste repository facility.

13. Development Milestones
2015
2016
2017
2018
2019
2020
2022
2026
2027
2031

256

Conceptual design completed
Start of basic engineering phase
Completion of CNSC pre-licensing Phase 1 Vendor Design Review
Commenced CNSC pre-licensing Phase 2 Vendor Design Review
Commenced irradiation test program on graphite
Strategic Innovation Fund (SIF) Award from Canadian Government
Basic engineering complete; planned - CNSC VDR Phase 2 completion
Planned - secure necessary licenses
Planned - commence construction of a first full-scale IMSR
Planned - first IMSR plant in-service

SSR-W (Moltex Energy, Canada)
All content included in this section has been provided by, and is reproduced with permission of, Moltex Energy, Canada.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
Moltex Energy, Canada
country of origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa

Static Fuelled Molten Salt Fast Reactor
Coolant is molten salt MgCl/NaCl
No moderator
750 / 300 continuous as baseload
750 / 900 as 8-hour peaking plant
Forced circulation
AtmosphericCore Inlet/Outlet Coolant
Temperature (oC)575/625

Fuel type/assembly array

Molten salt fuel within vented fuel tubes

Power conversion process

Molten salt storage tanks feed into steam
turbines
N/A (uses recycled spent fuel)N/A (on-load refuelling, 10 days)~100

Fuel enrichment (%)
Refuelling cycle (months)
Core Discharge Burnup(GWd/ton)
Reactivity control
mechanismApproach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)

Negative temperature reactivity coef
Boron carbide shutdown assemblies
Eliminate hazard, passive engineering,60
22 50014 / 750
Yes

Fuel cycle requirements /approach

Spent CANDU fuel is recycled through
WATSS to create SSR-W fuelMolten salt fuel in conventional fuelAssemblies. Thermal energy storage.
Very
low cost
conversion of spent LWR
Conceptual
designDistinguishing features
Design status

1. Introduction
The Stable Salt Reactor - Wasteburner (SSR-W) is unique in its use of molten salt fuel replacing solid pellets
in conventional fuel assemblies. This brings the major advantages of safe molten salts without the technicalhurdles of managing a mobile liquid fuel. The reactor is fuelled with very low purity, reactor-grade plutonium
recycled from stocks of spent uranium oxide fuel and produced by a low-cost process called WAste To Stable
Salt (WATSS). The reactor outputs its heat as a stream of molten chloride salts, which can be stored in large
volumes at low cost, making the reactor a low-cost peaking power plant rather than being restricted to baseload
operation. This same system permits the entire steam cycle to be identical to that of coal-fired power stations
and for it to be operated completely independently of the nuclear plant. The steam cycle is therefore not subject
to nuclear regulations.

2. Target Application
The SSR-W is designed for countries with significant stocks of spent nuclear fuel. The reactor burns all of thehigher actinide component of that fuel, leaving a waste stream of relatively short-lived fission products only,
as it can recycle actinides from its own spent fuel as well as from spent CANDU fuel. The fuel cost is expected
to be negative, net of the reduced liability cost for disposal of the original spent fuel. It is designed to be capableof economically efficient electrical power peaking, but with the reactor itself running at constant power. It
therefore fills the need of national power systems for a low-carbon complement to intermittent renewableenergy sources.

257

3. Design Philosophy
The entire design philosophy is to reduce plant costs by simplifying the design and eliminating hazards insteadof merely containing them. This is done by combining the safety and operational benefits of molten salts withthose of conventional reactor components. Risks to the public are practically eliminated by design, and not
merely contained.
The key features of the design are expected to achieve: virtual elimination of the possibility of the release of volatile radiotoxic materials in any conceivable
accident, terrorist act, or act of war; consequently, approval by regulators for deployment of the SSR-W on smaller sites with smaller
emergency planning zones;
 a capital cost as low as $1000/kW when operated as a peaking plant, which is cheaper than fossil fuels,
and without relying on subsidies; modular design, with a single 300 MWe reactor assembled from road-transportable, factory-produced
modules;
 a fuel assembly form that is compatible with IAEA safeguards procedures used in reactors today.

4. Main Design Features
Power Conversion
A molten salt to steam boiler is proposed to generate steam. The turbine is currently being specified for a 300
MWe plant similar to modern fossil fuel plants. There can be a larger turbine of up to 900 MWe installed, tobe coupled with the solar salt storage or multiple smaller turbines. This will depend on the local electricity gridneeds and economics.

Reactor Core
The reactor core’s horizontal cross section is a circle mostly filled concentric hexagonal rings. The innermost
rings of hexagons in the core are fuel assemblies, with shutdown assemblies spaced out along the periphery.
The gaps between these shutdown assemblies are filled in with permanent reflector and shield assemblies. The
gap between the outermost hexagonal ring and the circular tank wall is partially filled by the suction pipes anddowncomers of the six coolant loops. The remaining space is filled with a partial hexagonal ring of slots forspent fuel assemblies.
All the hexagonal assemblies are held in position on a diagrid structure at the bottom of the core and by thecollective interlocking arrangement of their support structures in the area above the coolant level, to form atank cover. Reactor coolant flows up the inside of the fuel assemblies, absorbing heat as it rises. Having risen
above the fuel assembly to the upper plenum, reactor coolant then flows via six loops to heat exchangers thatare outside of the reactor tank, before re-entering the tank in the volume underneath the diagrid.


Reactivity Control
No excess reactivity needs to be added to the core to compensate for fuel burn up because the combinationof frequent on power refuelling and high negative temperature reactivity coefficient allow the core to
generate constant power between refuelling steps. The small drop in reactivity is compensated by a small
fall in average fuel salt temperature.
No reactivity shims or control rods are required at any time under normal operating conditions, eliminating
the potential for control failures that can lead to an increase in the core reactivity.
Shutdown is achieved with shutdown assemblies. This is expected to control activity through radial
neutron leakage.

Reactor Vessel and Internals
The reactor coolant loops take heat from the fuelassemblies and transfer this heat to the thermal
storage system. This is achieved through six coolant
salt pumps, which feed six primary heat exchangers
by drawing coolant from the tank periphery throughthe suction pipes. This coolant is then discharged to
the lower plenum via the downcomers and jet
nozzles. The tube connections to the heatexchangers pass in angled paths through theconcrete biological shield, to prevent a shine path(neutron or gamma radiation escaping pastshielding).
In normal conditions, the flow is forced by pumps
to the lower plenum (below the diagrid) and sucked
SSR-W Coolant Loopfrom the upper plenum (coolant volume above theheight of fuel), while also having a portion of thecoolant bypass the suction pipes and flow freely down the tank wall to the lower plenum.
258

Fuel Characteristics
The fuel salt is a mixture of 45 mol% potassium chloride (KCl) and 55 mol% actinide trichlorides (UCl3,
PuCl3, etc.) and lanthanide trichlorides. It is redox stabilised to render it non-corrosive to steel by inclusion of
metal zirconium in each tube, which maintains the salt in a strongly reducing state incapable of dissolving
chromium from steel.

Fission Product Management
The volatile fission products traditionally of concern, such as cesium and iodine, are salts and remain trappedin the fuel. Noble gas radioactive isotopes are held up by the fuel pin vents, which hold it until the gas contains
acceptably low radiation levels. Those gases are then filtered by the reactor atmosphere conditioning system.

5. Safety Features
Engineered Safety System Approach and Configuration
The design philosophy is to follow the internationally accepted
principle of the risk mitigation pyramid. Therefore, the focus is touse inherent characteristics to eliminate hazards and to use
engineering safety features to provide additional confidence andbackup to the inherent physical characteristics.

Decay Heat Removal System
Natural convection of the primary coolant salt would continue inthe event of a reactor shutdown or pump failure. Heat would thenbe transferred to the tank walls, where a finned air duct toatmosphere around the tank walls would take decay heat away
indefinitely in an accident scenario.

Emergency Core Cooling System

Hierarchy of Controls

There is no safety injection system as the coolant has a very high boiling point and there is no credible leakage
path to allow the core to be uncovered.

Containment System
There are no significant internal pressures within the SSR-W. The primary containment for the reactor is thetube wall. The secondary containment is the coolant salt, which absorbs fission products in the event of tube
failure. The third is the tank itself. The fourth is the concrete structure. Above the tank is an argon space, which
has a stainless steel liner surrounded by a ~1 m concrete wall serving as the biological shield. The reactor
building walls are ~300 mm thick reinforced concrete. This serves as both a building structure and a shield.

Chemical Control
The reactor coolant salt chemistry is maintained in a reducing state through the use of an electrolytic cell (theredox control unit). The flow of coolant salt for the reactor coolant redox control unit is normally provided by
the leak-off flow that is generated in the reactor coolant loop vacuum break line when the reactor coolant
pumps are in normal operation.

Spent Fuel Cooling Safety Approach / System
Several positions along the periphery of the reactor allow the fuel assemblies to cool down after being removed
from the core. Afterward, they can be brought back into the WATSS process to allow them to be recycled into
new fuel for the reactor.

6. Plant Safety and Operational Performances
The design intent is that the release of radioactivity in excess of the agreed operating discharge authorization
is expected to be less than 10^-6 per year. The ramp rates of the plant will be driven by the steam side, not thenuclear side.

7. Instrumentation and Control System
There is a separate protection and monitoring system. The protection system is hardwired for a four-channel
coding system. The control system is a supervisory control and data acquisition (SCADA) system. The
monitoring system is a distributed data acquisition and monitoring system with standard monitoring software.

259

8. Plant Layout Arrangement
The reactor biological shield (RBS)
has a hexagonal cross section. The
RBS is built up on the reactor basemat
and both can effectively be consideredas different parts of the same civil
structure. There is a lining on theinside of the RBS Vessel that forms
part of the reactor containment. The
reactor tank is in the lower volume of
the RBS vessel; the upper volume
encloses the containment gas space
above the reactor coolant level andprovides a boundary for this gas
space. The reactor tank is enclosed by
the concrete RBS vessel, which
provides protection against radiationfrom the reactor core and coolant.

SSR-W Civil Structure

9. Testing Conducted for Design Verification and Validation
The SSR-W and WATSS design is sufficiently developed to support an R&D plan that is now being executed.
The novel componentry of the SSR-W includes the primary heat exchanger, the fuel assembly’s gas venting
facility, emergency heat removal system heat exchangers, and the fuel handling infrastructure. A mediumscale (tens of kgs of molten salt) thermal hydraulic loop is under construction and will demonstrate thephenomena that need to be accurately modelled by the appropriate software. A small-scale (<1 kg of moltensalt) experimental rig has been built and operated to provide the first validation data for heat transfer
coefficients necessary to modelling the thermal hydraulic performance of the Moltex core.

10. Design and Licensing Status
Moltex was selected by NB Power and the Government of New Brunswick to progress development of the
SSR-W in New Brunswick, Canada, with the goal of deploying its first reactor next to the Point Lepreau
Nuclear Generating Station. In preparation for this, the SSR-W design is being subjected to the CNSC Vendor
Design Review (VDR) process. The design has successfully completed Phase 1 of the VDR. A successful
Phase 2 review will demonstrate that there are no fundamental barriers to licensing the design in Canada.

11. Fuel Cycle Approach
Current CANDU reactors run on the once-through natural uranium fuel cycle, which involves a single pass ofthe fuel through the reactor. Moltex plans to produce the fuel for the 300 MWe SSR-W (SSR-W300) fromused CANDU fuel stocks using the WATSS process. WATSS separates out the fission products into a lowvolume high-level waste stream, produces a high-volume low-activity depleted uranium alloy and the SSRW300 fuel.

12. Waste Management and Disposal Plan
Moltex’s WATSS and SSR-W technology consume and destroy a very large proportion of the long-lived
higher actinides in spent CANDU fuel, while generating useful power. Other radionuclides present in CANDU
fuel are not consumed and remain as waste. Most of these radionuclides will be in a salt waste form that canbe safely disposed of in the proposed Canadian deep geological repository.

13. Development Milestones
2014
2014
2015

2017
2018
2018
2018
2020
2021
2021

260

UK patent granted for use of unpumped molten salt fuel in any reactor.
Independent capital cost estimate complete.
Pre- conceptual design complete and key claims validated by the UK’s
National Nuclear Laboratories.
Conceptual design completed and CNSC VDR commenced.
Master patent on static fuelled molten salt reactors granted in major
geographies. Several other patents progressing through PCT process.
Moltex successful in UK Government Advanced Modular Reactor
Competition.
Moltex signs agreement with NB Power to advance the development of the firstcommercial demonstration SSR-W in Canada.
Moltex signs ONWARDs agreement with Ontario Power Generation.
Moltex receives funding from the Government of Canada's Strategic Innovation Fund.
Moltex completes CNSC VDR Phase 1.

smTMSR-400 (CAS/SINAP, China)
All content included in this section has been provided by, and is reproduced with permission of, CAS/SINAP, China.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country of
CAS/SINAP, China
originReactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Molten salt reactor
LiF-BeF2-ZrF4-ThF4-UF4 fuel salt / Graphite
400 / 168

Primary circulation
NSSS Operating Pressure(primary/secondary), MPa

Forced circulation
Near ambientCore Inlet/Outlet Coolant
Temperature (oC)

650 / 700

Fuel type/assembly array
Power conversion process
Fuel enrichment (%)
Refuelling cycle (months)
Core Discharge Burnup (GWd/ton)

Molten salt fuelhelium/air/CO2 Brayton cycle
19.75
120 (batch-reprocessing off-line)~300

Reactivity control mechanism

Control rods, Online fuel addition, drain off
fuel salt
Passive60
40 000~10 / 3.8120
0.3g
Th+U loading initial, LEU addition online, FP
gas removal online, Batch-reprocessing offline

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements / approachDistinguishing features

Design status

Replaceable Reactor module (~10 years);
Passive safety Near 40% power contributed bythorium
Pre-conceptual design

1. Introduction
A three-step development route has been proposed by Shanghai Institute of Applied Physics, Chinese Academyof Science, (SINAP, CAS), to realize thorium-uranium breeding in molten salt reactor by the middle of thiscentury. smTMSR-400 is a 400MW(t) / 168 MW(e) small modular Thorium Molten Salt Demonstration
Reactor that forms part of the third step. It will demonstrate large-scale power produced based on the thorium
fuel cycle and verify the conversion properties. Most off-line pyro-processing techniques, such as, fluoride
volatility, vacuum distillation and electrochemical reduction, will be used in fuel cycle of smTMSR-400.
Demonstration at this stage focuses on the R&D efforts and can reduce the risk and cost. The option also enable
the gradual establishment of the thorium fuel and MSR supply chain system. smTMSR-400 will demonstrate
the highly intrinsic safety and engineering reliability of molten salt reactor, such as shutdown by negative
temperature feedback and draining off fuel salt, passive residual heat removal systems. The aim of safety
design is to avoid any off-site emergencies and therefore increase the siting flexibility. smTMSR-400 uses
small and modular designs to reduce and simplify the R&D challenge and difficulty of MSR. Modular design
will make MSR more flexible to deploy at different places with various demands. Modular manufacture andassembly can improve the quality of equipment and enhance safety and economy. Modular construction canalso reduce the financing cost and risk.

2. Target Application
smTMSR-400 is designed as a thorium convertor and in situ burner driven by low enriched uranium. The fuelcost is expected to be cheaper than pure uranium fuel mode. After offline batch reprocessing, thorium and261

uranium can be recycled into a new reactor in order to minimum the spent fuel mass as well as enhance theneutron economy.
It will be applied as a high temperature heat source, which not only can be used for electricity generation, butalso can satisfy the energy diversified demands, such as, seawater desalination, heat supply, supercritical steam
supply for industry demands and hydrogen production, etc.

3. Design Philosophy
The design philosophy of smTMSR-400 includes:- Using mature technologies and experiences accumulated from TMSR-LF1 project and potential
technologies to be developed in the next years.
- The primary loop is designed as a compact reactor module and can be replaced every 8-10 years to solve
the long-term irradiation problems of materials.
- -Online fuel addition and offline batch reprocessing mode is used. The power contribution from thorium
should be more than 40%. The spent fuel can be used for offline pyro-processing and recycled for threetimes.
- -Passive safety design, consist of negative temperature feedback, passive fuel salt discharge, passive residual
heat removal systems, and radioactive nuclides retention by salt coagulation.
- Heat storage system is applied for comprehensive energy utilization and peak-shift of electrical demand.

4. Main Design Features
Power Conversion
The heat can be applied for electricity generation by helium/air/CO2 Brayton power system based on theirtechnological maturity and application scenarios. Also, it can be applied for seawater desalination, heat supply
and steam supply.

Reactor Core
Reactor module is designed as a compact loop structure with reactor core, three salt/salt heat exchangers,
one centrifugal pump and connecting pipelines. Reactor core is filled with LiF-BeF2-ZrF4-ThF4-UF4 fuelsalt and hundreds of prismatic graphite blocks. 235U is 19.75% enriched. Three salt/salt plate heat exchangers
are arranged upon the reactor core side by side for compact layout. The size of reactor module is about ~3.8m in diameter and ~10 m in height, which is suitable for railway transportation. Reactor module is packed in asafety vessel for additional radioactive confinement barrier. The reactor vessel and metal internals in touchwith fuel salt are made by nickel-based alloy.
The fuel salt is heated up to 700°C by fission energy in the active core (fuel salt channels in graphite
moderated core), then flows through the upper plenum, upper pipeline, driven by the pump, and then to thethree plate heat exchangers. After heat transfer, the fuel salt is cooled down to 650°C, then flow through thedown-comer, lower plenum and back to the fuel salt channel.

Reactivity Control
Six control rods are used to control the reactivity change covering from subcritical status, power lift
condition, xenon poison, operation disturbance, and excess reactivity. Most of control rods are used to adjust
the power lift condition due to the negative temperature feedback.

262

Reactor Vessel and Internals
The reactor vessel is made of nickel base alloy Hastelloy-N, which shows good mechanical behaviour andcorrosion resistance under high temperature. The internals include the support plates of core, the barrel, thearc plate of the upper and lower chamber, the tubes of control rods, and so on. Composite material will be also
used in smTMSR-400.

Fuel Characteristics
The fuel salt consists of LiF-BeF2-ZrF4-ThF4-UF4. LiF-BeF2 is used to lower the melting point and keep agood neutronic and thermal-hydraulic properties. In order to reduce the neutron absorption and tritium
production, Li-7 is required to be enriched more than 99.99%. ZrF4 is used to prevent the precipitation of
uranium and thorium in the case of accidental introduction of oxygen impurities. UF4 with 19.75% enrichment
is the start-up fuel, and ThF4 is the breeding fuel.

Fission Product Management
During the operation, the fission gases are online swept through bubbling method to the off gas system, where
they will decay to a low radioactive level and then emit into the atmosphere. After 10 year operation, soluble
fission products will be discharged into the drain tank and decay for several years, then will be sent to theoffline pyro-processing facility for separation and permanent disposal.

5. Safety Features
Engineered Safety System Approach and Configuration
smTMSR-400 contains two sets of control rod system with different driving mechanisms, which willautomatically go down in accidents. Low excess reactivity will also make it possible to shut down with thenegative temperature feedback alone. After that, fuel salt will be drained off for long-term shutdown.

Decay Heat Removal System
smTMSR-400 is designed with two passive decay heat removal systems (DHRS). Natural circulation will formin the reactor module in case of force circulation failure. When the nuclear heat supply system fails, two kindsof DHRSs will ensure long-term cooling of fuel salt. One is located around the safety vessel in the silo
(DHRS1), another is in the fuel salt drain tank directly installed below the reactor vessel (DHRS2). DHRS1 is
operated under normal accidents to cool the safety vessel and reactor vessel. DHRS2 is triggered after thefreeze-valve opens passively by increased accident temperatures or by active heating.

Containment System
There are four containment barriers in smTMSR-400. The first one is the reactor module made by hightemperature-resistant and corrosion-resistant nickel-based alloy, the freeze valve function prevents the firstbarrier from exceeding its temperature limits and thus to keep its integrity under accidents. The second one is
the safety vessel, which will ensure gas tightness and contain the leaked salt under beyond design basic
accidents. The third one is the fuel salt itself, which has a large degree of retention on some importantradioactive elements. The solidification of fuel salt will prevent further leakage. The final one is theunderground construction, which can effectively prevent the spread of radioactive materials in an accident andcan resist natural disasters and terrorist attacks.

Chemical Control
Online chemical control system with addition of metal Be or ZrF4 is designed to control the oxidation-reduction
potential of fuel salt.

Spent Fuel Cooling Safety Approach / System-The spent fuel salt will stay in the drain tank for several years with continuously cooling.

6. Plant Safety and Operational Performances
There is no core meltdown accident in smTMSR-400, the fuel salt itself is at liquid state during the normal
operation. There are no significant safety risks, such as high pressure, violent chemical reaction or explosion,
which may lead to the large early release of radionuclides. The high temperature margin, the large thermal
capacity of fuel salt and graphite, the passive decay heat removal ability and the low excess reactivity guarantee
the integrity of containment boundary. Potential less leak of gas or salt in the case of pipe or valve corrosion
can be limited by multi-physical barriers. It is expected the eliminate the possibility of massive release ofradioactivity. The capacity factor can reach 95% due to the ability of online fuel loading.

7. Plant Layout Arrangement
The reactor module, fuel management and gas removal system are under the ground. The heat storage systemand heat utilization system are above the ground. Power expansion can be realized by employing multiple
reactor units. The power plant can adopt single unit, two units and different units at one site, and realizes
different power output to meet the energy needs of different application sites.
263

8. Testing Conducted for Design Verification and Validation
To be developed.

9. Design and Licensing Status
The design is in a pre-conceptual stage of development and licensing activities have not yet been undertaken.

10. Fuel Cycle Approach
The fuel cycle in smTMSR-400 belongs to a semi-enclosed type. During the operation, fuel will feed online
and some of fission gases are removed and treated in the off-gas system. After each 10 year, the spent fuel salt
will be transported to the pyro-processing facility for recovering the uranium, thorium and carrier salt. The
remained TRU and FPs will be minimized and be permanently disposed.

11. Waste Management and Disposal Plan
A pilot stage pyro-processing facility will be established in 2030s.The spent fuel in smTMSR-400 will be sent
to the facility for fuel salt recovery and to minimize the radioactive inventory for disposal.

12. Development Milestones
2018-2021
2022-2024
2025-2027
2028-

264

Pre-conceptual design phase and technology research.
Conceptual design phase and technology validation
Engineering design phase.
Projected deployment (start of construction) time.

Compact Molten Salt Reactor (Seaborg
Technologies, Denmark)All content included in this section has been provided by, and is reproduced with permission of, Seaborg Technologies, Denmark.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
Seaborg Technologies ApS, Denmark
country of origin
Reactor type
Coolant/moderator
Thermal/electrical
capacity, MW(t)/MW(e)CMSR POWER BARGE

Molten salt reactor / thermal spectrum
Fluoride fuel salt / NaOH salt (patented
moderator)
250 / 100 per CMSR

Primary circulation
Operating Pressure(MPa)(primary/secondary/NSSS)

Forced circulation
0.45 / 1.05 / 18

Core Inlet/Outlet Coolant
Temperature (oC)

600 / 670

Fuel type/assembly array
Power conversion process

HALEU / Molten salt fuelSuperheated steam driven turbine /generator (Rankine Cycle)

Fuel enrichment (%)
Refuelling cycle (months)

HALEU/different scenarios considered
No online refuelling(One Fuel Cycle: 144 months / 12
years)
Pending Fuel Enrichment Selection

Core Discharge Burnup(GWd/ton)
Reactivity control
mechanismApproach to safetysystems
Design life (years)
Plant footprint (m2)

Negative temperature coefficients,
regulating and safety rods, fuel salt
draining
Automatic Shutdown, Passive Cooling
12 per CMSR / 24 per Power Barge
200 - 800 MWe – 5 000 – 14 000 m2
Depending on the Power Barge model.

RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements /approach

5.5 / 2.5
50 – 60
Under review
HALEU one cycle
Off-Site reprocessing

Distinguishing featuresCMSR integrated into a floating nonself-propelled Power Barge; new
CMSR units installed after 12 years.
CMSR Power Barge operates for 24
years. Moduar Nature provides output
from 200-800 MWe
Liquid fuel which is the primarycoolant, and a liquid moderator.
Conceptual design

CMSR
Design status

1. Introduction
The Seaborg Compact Molten Salt Reactor (CMSR) is an advanced, small, and modular molten salt reactor
characterized by using a liquid fluoride molten salt fuel and a liquid patented molten hydroxide moderator.
The CMSR produces approximately 250 MWth (MW-thermal) through the fission of High-Assay LowEnriched Uranium (HALEU) in a thermal neutron spectrum. The liquid moderator, which does not accumulate
irradiation damage, enables the CMSR an operating life of 12 years, during which no refuelling is required.
The fuel and moderator salts have high melting points. As a result, the CMSR needs no pressurization above
what is needed to circulate the salts. The CMSR has unique inherent safety features. The CMSR is deployedon modular and standardized non-self-propelled Power Barges, scalable from 200 MWe to 800MWe, and with
an operating life of 24 years.
2652. Target Application
The CMSR Power Barge is designed to generate power for the electrical grid and for industrial applications
such as hydrogen electrolysis. Furthermore, the CMSR Power Barge can supply high temperature heat for nonelectrical applications such as district heating, water desalination, and industrial processes.

3. Design Philosophy
The CMSR design philosophy is focussed on implementing inherent and passive safety features as well as
fully automated shutdown. The design ensures high proliferation resistance. The design is aimed to reducecomplexity and ensure increased manufacturability and modularity. Proven technologies are used as far as
possible and innovative technologies are applied only where necessary to facilitate the design goals. The
modular and compact nature facilitates the integration of the CMSR into the CMSR Power Barge. The
philosophy of the CMSR Power Barge is to combine the CMSR design with the construction and high-quality,
matured capabilities of shipyards, thus providing extensive scalability.

4. Main Design Features
Power Conversion
Each CMSR liquid fuel fluoride molten salt loop transfers heat generated in the core to a secondary liquid
heat-transfer fluoride salt via the primary heat exchanger. The secondary salt circuit then transfers heat to thePower Conversion system via a steam generator, where live superheated steam at up to 18 MPa and 565ºC is
produced. The steam produced from the two generators drives a single conventional condensing steam turbinewith exhaust positioned axially into the seawater-cooled condenser. Electrical output from the connected
generator allows delivery of 200MWe net from each Power Module through a common electrical interface tothe high-voltage onshore electrical transmission grid. Through this arrangement, the Power Barges can provide
between 200 and 800MWe net to the grid depending on the number of modules installed.

Reactor Core
The reactor core of the CMSR consists of an upper plenum, lower plenum, and fuel tubes which contain andguide the molten fuel salt through the active core, whilst also maintaining separation between the fuel salt andmoderator salt. The CMSR consists of 250 fuel tubes that are arranged in a 60° triangular tube pitch patternwith control rod guide tubes inserted in between. Most of the heat is produced in the fuel salt and is the primaryheat carrier from the active core to the heat exchanger.

Reactivity Control
In the CMSR design there are two types of control rods used: Regulating Rods for compensating for shortterm reactivity changes, and Safety Rods, for extinguishing the nuclear chain reaction promptly when needed.
The design also contains a highly reliable, redundant and supplementary method for extinguishing the nuclear
chain reaction achieved through draining of the Fuel Salt from the reactor to a Fuel Salt Drain tank.

Reactor Vessel and Internals
The reactor vessel of the CMSR has a height of 5.5 meters and a diameter of 2.5 meters excluding shielding
and thermal insulation, the reactor vessel lid has the penetration for the control rods. Inside the reactor vessel,
the 2m active core contains the fuel salt in the tubes.

Fuel Characteristics
The fissile component of the fuel is HALEU. The enrichment level is chosen to enable an operating lifetimeof 12 years without refuelling. The fuel consists of NaF, KF, and UF4, with partial reduction of UF4 to UF3
to limit degradation of the structural materials. Partial reduction is maintained throughout the operating lifetimeby the chemical control system. Most fission products form soluble fluoride species which are retained in thefuel for the lifetime of operation. A fraction of these fission products is insoluble, mostly as noble metals andvolatile species such as noble gases. Noble metals are isolated within the Fuel Salt System. Volatile species
are continuously removed from the primary circuit during operation through the Fuel Salt Off-gas System; thisstream is treated and conditioning through different operation units.

5. Safety Features
Engineered Safety System Approach and Configuration
The approach and configuration of engineering safety system relies on a tailored defence in depth, with a topdown focus on the three fundamental safety functions. In addition to the engineering systems, the inherent
features of the CMSR are considered with respect to these three fundamental safety functions, assessed, andcascaded down at the defence in depth and plant level safety architecture. The inherent features considered in266

the conceptual phase and supported by R&D but not limited, are: 1.- Control of reactivity; a) use of gravity toreach a subcritical configuration, b) negative temperature coefficient, c) moderator. 2.- Cool by removal ofheat from the reactor and from the fuel/waste storage; a) high melting point, b) volumetric heat capacity, c)passive systems. 3.- Confinement of radioactive material, shielding against radiation and control of planned
radioactive releases, as well as limitation of accidental Retention of fission products; a) atmospheric pressure
operation, b) hydrogen risk elimination.

Decay Heat Removal System
The CMSR Power Barge has a main line and a diverse line for decay heat removal. Both rely on the principleof draining the fuel salt from the reactor and primary circuit into a subcritical geometry fuel salt drain tankusing gravity and other passive means. Once the fuel salt is in the drain tanks, the decay heat can be removed
through the main active line. If this main active line fails or during a Station Black-out (SBO), a diverse passive
line is required to extract the decay heat. This system enables cooling for 72h without active systems/external
action.

Emergency Core Cooling System
Considering the atmospheric operating conditions and ability to drain the fuel salts, the concept of ECCS iscovered and implemented via the decay heat removal systems.

Containment System
A containment enclosure system (known as the “safety enclosure” in the marine industry) prevents releases tothe atmosphere. The containment enclosure is a leak-tight structure and comprises of non-accessible (reactor
and drain tank compartments) and accessible compartments (equipment compartments).

Chemical Control System
Chemical control systems are implemented for each of the molten salt circuits. The fuel salt chemical control
system maintains partial reduction of UF4 to UF3 and continuously removes fission products from the primarycircuit to maintain the redox potential that is necessary to minimize uniform corrosion during normal operationand transients. The moderator salt chemical control system limits corrosion in the moderator circuit using
proprietary technology. The secondary salt chemical control system ensures that the secondary coolant
chemistry remains within operational design limits.

Spent Fuel Cooling Safety Approach / System
Each reactor in the CMSR Power Barge operates 12 years without refuelling with the drain tanks described in
section 5-b fulfilling the role of a spent fuel cooling system after the 12 years of operation and until
decommissioning. The fuel salt is allowed to solidify in the drain tanks and passive means of decay heatremoval are employed in a similar fashion as dry cask storage for conventional nuclear spent fuel.

6. Plant Safety and Operational Performances
The unique nature of the molten fuel salt drastically reduces the Core Damage Frequency and Large Early
Release Frequency. The high boiling point of the molten salt does not require the high pressures typicallyassociated with LWRs. The on-line noble fission products are removed via helium carrier gas and purification.
Additionally, as the fuel cools and solidifies, non-noble fission products are retained in the salt, minimizing
the release to the environment. Experiments are currently being designed to validate these claims. The 12-year
CMSR lifecycle reduces the Aging Management requirements and presents an opportunity to implement a
novel maintenance approach. Based on the principles of Reliability Centred Maintenance, efficient schedulingof maintenance activities allows a single CMSR to achieve a Capacity Factor of 90% including a reserve ofseveral weeks for unplanned outages. The 24 years cycle of operation is achieved with the 2nd set of CMSRs.

7. Instrumentation and Control System
The Seaborg CMSR Power Barge I&C systems provide automated control, monitoring, and protection
functions of the CMSR Power Barge. The Seaborg CMSR I&C architecture reflects the modular nature of the
CMSR Power Barge and is designed to ensure separation of safety and non-safety systems. Each CMSR
Reactor shall have its own protection system which shall monitor critical parameters and if required shall
automatically initiate reactor trip and relevant safety functions to achieve and maintain the CMSR systems ina safe state.

8. Plant Layout Arrangement
The CMSR Power Barge consists of one or more power modules connected to an aft end module(accommodation, offices, conference rooms, main control room, supplementary control room, emergency
power, and barge utilities) and a fore end module, where the High Voltage substation shall be positioned. The
entire hull is designed with a complete double hull, which protects the safety enclosure, the CMSR auxiliary
systems, and the emergency diesel generators, diesel tanks, and switchboards. The CMSR Power Barge is not
self-propelled and shall be equipped with suitable towing and mooring facilities.

267

200MWe Power Module of the CMSR Power Barge

Cross-sectional view

9. Testing Conducted for Design Verification and Validation
In house laboratories verify corrosion control processes in the fuel and moderator salts, including flow loops
to enable properties to be verified in a relevant environment. Flow loops and additional custom designed
apparatus are used to validate simulation methodologies such as thermal hydraulics. Partnerships establishedwith major institutions enhance verification and validation capabilities, both experimental and modelling.
Collaborative projects established to determine: a) physicochemical properties of the fuel salt, b) thermal
scattering kernels measured for solid and those for molten NaOH. Irradiation experiments for verification andvalidation of materials and salts properties are also being established. In addition, in-house robust capabilities
to model and simulate in parallel the design are constantly enhanced.

10. Design and Licensing Status
The CMSR is at a conceptual design stage with an emphasis on a research and development approach throughthe development of test facilities and loops to support licensing of a first of a kind CMSR, simplify design andreduce regulatory risk. The licensing of a first of a kind CMSR focuses on a standard design assessment basedon a technology-inclusive international framework and complemented by the ongoing development of riskinformed performance-based approaches. Preliminary engagement is being evaluated with several national
regulators in parallel to support international deployment of the CMSR.

11. Fuel Cycle Approach
Strict atmosphere control is always maintained once the fuel has been manufactured to prevent the ingress of
impurities such as oxygen and moisture. Fuel is loaded into the drain tanks prior to operation. From here, it isheated to the operating temperature and transferred to the primary circuit. The fuel is enriched to a level whichenables operation for 12 years without refuelling. At the end of lifetime, the fuel is returned to the drain tanks
where they are cooled. After 24 years of operation, the Power Barge is returned to a specialized dedicateddecommissioning facility where further decay heat removal of the drain tanks is ensured followed byreprocessing of the fuel.

12. Waste Management and Disposal Plan
Waste management and disposal approach is flexible and can be adapted to fit national polices of the host stateand contractual arrangements with a “greenfield” philosophy. Industry practices are applied to manage andclassify waste with higher level wastes (CMSR, spent fuel prior to reprocessing, etc.) predominantly aimed forstorage/disposal in dedicated facilities abroad with reprocessing of the spent fuel favored, and lower-level
wastes aimed for storage/disposal in the host state country or abroad provided adequate additional provisions.

13. Development Milestones
2020
2022
2024
2026
2028

268

CMSR Feasibility Statement received by ABS
Barge Approval in Principle by ABS
End of Concept Verification
End of Design
Delivery of first BargeCopenhagen Atomics Waste Burner(Copenhagen Atomics, Denmark)All content included in this section has been provided by, and is reproduced with permission of, Copenhagen Atomics, Denmark.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
Copenhagen Atomics, Denmarkof origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Molten salt reactor
Fuel salt/heavy water
100 / NaN

Primary circulation
NSSS Operating Pressure(primary/secondary), MPa

Forced circulation
0.05 – 0.25 / 0.1 – 0.25

Core Inlet/Outlet Coolant
Temperature (oC)

600 / 650 - 700

Fuel type/assembly array
Power conversion process
Fuel enrichment (%)
Refuelling cycle (months)
Core Discharge Burnup(GWd/ton)

7LiF-ThF -(TRU)F / none
4
3

Reactivity control mechanism
Approach to safety systems
Design life (years)

Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle requirements /approach

Heat source
TRU or RGPu
NaN
900 – 1000
Heavy water level adjustment
Passive
5 years for the reactor vessel,
minimum 50 years for thesurrounding building, and infinite
life for the salts and heavy water.
2 500
12 / 2.4
40
NaN
Spent fuel initiated / conversion to
Th-U cycle

Distinguishing features

Liquid moderator, Low fissile
inventory, and Potential for breeding

Design status

Detailed design / Equipment
manufacturing in progress

1. Introduction
The Copenhagen Atomics Waste Burner is an autonomous small single module 100 MW(t) heavy watermoderated, fluoride salt based, thermal spectrum, molten salt reactor the size of a 40 feet shipping container.
The reactor doesn’t require refuelling, human intervention, or maintenance for its 5-year design life andoperates completely autonomously with passive decay heat removal. The salts and heavy moderator have, in
principle, infinite lifetime and will be reused in subsequent reactors, while the reactor vessel is the consumableand fission products are left as a by-product.
The Copenhagen Atomics Waste Burner can burn transuranics and has the potential to transition to breeding
within 3 years of operation. The core and liquid moderator, fuel salt loop, blanket salt loop, coolant salt loop,
fission product separation systems, dump tank, heat exchangers, and pumps, are all contained in a leak-tightsteel containment, the size of a 40 feet shipping container. In this way the Copenhagen Atomics Waste Burner
employs three barriers to nature.

2. Target Application
Copenhagen Atomics will design, licence, build, operate, and decommission all Copenhagen Atomics Waste
Burners. Each reactor delivers heat as a service, in the form of up to 560°C nitrate salt, to the customer. In
order to facilitate early mass deployment, Copenhagen Atomics plans to initially sell to customers that needdouble digit GW thermal capacity to produce liquid products, such as ammonia production or desalination ofwater.
269

3. Design Philosophy

Copenhagen Atomics see autonomous operation without maintenance and road transportable completely
factory assembled single unit reactors as essential for mass deployment and scaling of nuclear power toterawatt levels. To this end Copenhagen Atomics are developing thermal spectrum reactors with high specificfissile inventory (thermal power to total fissile inventory) and online fission product separation and targetingburning of spent nuclear fuel transuranics, kickstarting a thermal spectrum thorium breeding cycle.
Copenhagen Atomics is focused on building and testing ‘minimal viable product’ iterations of our Copenhagen
Atomics Waste Burners before going through a commercial licensing process. As part of this development
process Copenhagen Atomics have developed most of the components that go into a reactor from scratch andwith the ability to operate for 5 years without need of maintenance. This includes canned active
electromagnetic pumps, electrochemical sensors, autonomous consensus steering electronics and software, anda lot of the infrastructure needed for operation such as salt purification and data acquisition.

4. Main Design Features(a) Power Conversion Unit
The Copenhagen Atomics Waste Burner delivers heat as a service and is operated by Copenhagen Atomics.
The customer of the heat can choose to couple the reactor heat output with a power conversion system.

(b) Reactor Core
The Copenhagen Atomics Waste Burner uses 7LiF-ThF4-(TRU)F3 or 7LiF-ThF4-RGPuF3 kickstarter fuel salt,
7
LiF-ThF4 blanket salt, and unpressurized room temperature heavy water moderator. Bred uranium from theblanket is transferred to the fuel salt online.
The heavy water moderator is circulated and cooled, and thermally isolated from the molten salts, so that thevast majority of moderator heating is through the thermalization of neutrons. The moderator cooling isachieved with an external commercial air conditioner, consuming a small fraction of the produced heat tomaintain the moderator temperature. Employing a blanket salt allows for a compact core while maintaining a
low neutron leakage. The reactor core design enables self-sustained breeding in thermal spectrum with thethorium fuel cycle, if fission products are removed online and composite core construction material isemployed. Current versions of the Copenhagen Atomics Waste Burner prototype’s core uses stainless steel
core construction material and fission product separation limited to volatile gases.

(c) Reactivity Control
During normal operation the liquid moderator allows core reactivity control through a simple heavy water leveladjustment. Gradual lowering of fissile inventory from transuranic burning period can be compensated by
gradually increasing the heavy water level. Gradual build-up of fissile inventory from breeding when
predominantly running on thorium cycle can be compensated by gradually lowering the heavy water level.
The highly negative temperature reactivity feedback of the Copenhagen Atomics Waste Burner is relied upon
as a second and independent reactivity control mechanism.
The fuel salt, blanket salt, and heavy water moderator are all respectively being continuously and passively
drained from the core at a high rate and actively pumped and cooled before going back into the core. If poweris shut off or any of the pumps tripped the reactor will shut down, due to the lack of fuel, moderator, and orreflector blanket.

5. Safety Features(a) Engineered Safety System Approach and Configuration
The entire reactor operation and shutdown is designed to be completely autonomous with decay heat removed
passively. Safety is ensured through inherent safety instead of operator action.

(b) Decay Heat Removal System and Emergency Core Cooling System
The fuel salt is pumped through the dump tank so that in the case of a pump trip the fuel salt drains quickly tothe dump tank, where the decay can conduct through the tank bottom and through the primary radiation
shielding to the surroundings. This enables completely passive decay heat removal but also means that there’s
always a parasitic heat loss through the same heat path during normal operation. This is an economic trade-off
for enhanced passive safety.

(c) Containment System
The reactor containment consists of three physical barriers to nature. The first barrier is the fuel salt and fission
product trap wetted structure and components. The second barrier is the leak tight 40 feet size vessel thatsurrounds the first barrier, reactor core, reactor coolant salt, and other components. The third barrier is anexterior onsite constructed leak tight containment that also serves as radiation shielding. Another path isthrough the reactor coolant and secondary coolant salt which is served to the customer. Here the three heatexchangers serve as three physical barriers, first the fuel salt to reactor coolant salt heat exchanger, second the270reactor coolant salt to secondary coolant salt, both of these are located inside the 40 feet sized vessel, and third
the heat exchanger or boiler on the customer side.

(d) Spent Fuel Cooling Safety Approach / System
If online fission product removal is employed then the fuel salt can be transferred to a new reactor unit, at theend of life or after failure of one unit. If only online volatile fission product removal is employed thenintermediate salt cleaning will be necessary before transfer to a new unit. Thus, due to the continuous use of
the fuel there is no spent fuel. Fission product traps are envisioned to be left in the used reactor vessel, where
their decay heat can be passively removed through the normal decay heat removal system for a cool down
period before decommissioning. In the case of a failure of the salt wetted barrier where the fuel salt can’t betransferred through the normal means then it would be left for a cool down period before decommissioning,where the fuel would be separated, cleaned, and reused.

6. Plant Safety and Operational Performances
The Copenhagen Atomics Waste Burner’s safety is not dependent on the use of the heat that is delivered to thecustomer. The reactor doesn’t require refuelling, human intervention, or maintenance for its 5-year design life
and operates completely autonomously with passive decay heat removal. A plant is envisaged to have dozens
of reactor units, placed together and inaccessible during operation, besides for swapping of reactor vessels
under observation. Satisfactory safety statistics for commercial operation will be determined from reactor
prototype operation.

7. Instrumentation and Control System
Copenhagen Atomics makes/writes all of our own electronics and software for the Copenhagen Atomics Waste
Burner. The system uses redundant consumer grade electronics running consensus steering for autonomous
operation of the reactor throughout its 5-year lifetime. All instrumentation and control systems are sealed inside
the 40 feet sized reactor vessel and can’t be accessed or altered after production.

8. Plant Layout Arrangement
The reactor and containment structures function as an adjacent heat source for existing or new industrial plants
on site but are inaccessible from the rest of the site with a land claim of roughly 2500m2 per 100MW(t) reactor.
Since the customer is served up to 560°C molten nitrate salt the reactor site and user can also be physicallyseparated by up to kilometres with pipelines in between.

9. Testing Conducted for Design Verification and Validation
Copenhagen Atomics relies extensively on prototype testing and is envisaging testing several prototype
reactors before offering a commercial reactor.
Currently, Copenhagen Atomics are operating two dozen pumped molten salt loops, four dozen static molten
salt test systems, including some with fertile LiF-ThF4 salt and some with LiF-ThF4-PuF3 salt test slated for2022-2023, two 1MW non-fission prototype reactors are under construction to run with water and FLiNaK
respectively, and multi ton scale fertile salt production is under construction. Both pumped molten salt loops
with fuel salt and 1MW non-fission prototype reactors with fertile salt are slated for 2023-2024.

10. Design and Licensing Status
Copenhagen Atomics has started licensing for production of multi digit tonnes of thorium and natural uranium
for testing of non-fission prototype reactors. Copenhagen Atomics is interacting with nuclear safety authorities
on testing of a 1MW prototype reactor. Copenhagen Atomics does not intend to engage in a commercial
licensing process for a 100MW(t) reactor before a prototype reactor has been tested.

11. Fuel Cycle Approach
The Copenhagen Atomics Waste Burner uses 7LiF-ThF4-(TRU)F3 or 7LiF-ThF4-RGPuF3 fuel salt with 3% mol(TRU)F3 or RGPuF3 kickstarter fuel salt and 7LiF-ThF4 blanket salt. The reactor can transition to a breeder
reactor using the heavy water level for coarse reactivity adjustment. The reactor core design enables selfsustained breeding in the thermal spectrum and thorium fuel cycle if fission products are removed online andC/C composite core construction material is employed. The Copenhagen Atomics Waste Burner needs an initialload of roughly 100-200 kg of TRU or RGPu and doesn’t require any refuelling for its 5-year operation.
The Copenhagen Atomics Waste Burner design can also start on roughly 1 ton 5% enriched uranium using a
7
LiF-UF4 fuel salt and 7LiF-ThF4 blanket salt, and can achieve a conversion factor of around one within three
years of operation. After 5-year operation the burnup is roughly 200 GWd/tU and most of the fissile uranium
in the fuel salt is 233U that was produced in the blanket. The fuel and blanket salt can be reused in subsequent
reactors without adding makeup fuel.

12. Waste Management and Disposal Plan
The Copenhagen Atomics Waste Burner can transition to breeding within the first three years of operations
with online removal of fission products and C/C composite core construction material. However, the highpower density of the small core limits the operation life of the reactor due to neutron irradiation, but the salts271

and heavy moderator have, in principle, infinite lifetime and will be reused in subsequent reactors, while thereactor vessel is the consumable and fission products are left as a by-product.
After operation the reactor vessel and fission product trap are left to cool down before decommissioning. For
decommissioning it’s envisaged to separate fission products for vitrification or other uses and remelting of thevessel for non-nuclear reuse with low remaining activity and non-metals separated in the slag.

13. Development Milestones
2015
2015
2016
2017
2018
2019
2020
2022
2025
2028

272

Copenhagen Atomics was founded in 2015 by a group of passionate engineers andscientists meeting up since 2013 and based on an open-source model, where results andfindings are shared with the thorium molten salt reactor community.
First simulations of neutron economy and online fission product removal.
First static molten salt test.
First pressure driven circulation molten salt loop.
First pumped circulation molten salt loop.
Award two Danish research grants and start of collaboration with Alfa Laval.
First public funding round and first fertile salt test.
Completion of non-fission prototype of a 1MW(t) demonstration reactor.
First test of a 1 MW(t) demonstration reactor.
First test of a 100 MW(t) commercial reactor.

FUJI (International Thorium Molten-Salt
Forum, Japan)
All content included in this section has been provided by, and is reproduced with permission of, ITMSF, Japan.

Parameter

MSR-FUJI cut model

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

ITMSF, Japan

Coolant/moderator

Molten fluoride/graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

450 / 200
Forced circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

0.5 / 0.5

Number of fuel assemblies
in the core
Fuel enrichment (%)

N/ACore Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

Schematic diagram of MSR-FUJI

Molten salt reactor

565 / 704
Molten salt with Th and U

2.0 (0.24%233U + 12.0%Th). Pu or LEU
can be used
No mechanical limit for burnup
Continuous operation possible

Reactivity control
mechanismApproach to safety systems

Control rod, or pump speed, or fuelconcentration
Passive

Design life (years)

30

Plant footprint (m2)

<5000 (RB+SGB+TGB)

RPV height/diameter (m)5.40 / 5.34 (inner)

RPV weight (metric ton)

60 (made of Hastelloy N)

Seismic Design (SSE)Same as LWRs

Fuel cycle requirements /ApproachSelf-sustaining at FUJI-U3.
No online reprocessing, except gaseous
FP removal.
Spent fuel salt is reprocessed
Self-sustaining at FUJI-U3.
Spent fuel salt is reprocessed at off-sitefacility.
3 experimental MSRs were built.
Detailed design not started

Distinguishing features

Design status

1. Introduction
The Molten Salt Reactor (MSR) was originally developed at Oak Ridge National Laboratory (ORNL) in 1960s,
and three experimental MSRs were constructed. One of them was operated for 4 years without severe
problems. Thus, it is verified that the MSR technology is feasible. MSR-FUJI was developed since the 1980s
by a Japanese group (International Thorium Molten-Salt Forum: ITMSF), based on the ORNL’s results to
deploy it in the world.
Molten salt is stable and inert at high temperature and can be used at very low pressure. Since core meltdown
or steam/hydrogen explosion is impossible, high safety can be achieved. MSR-FUJI is size-flexible as from 25
MW(e) to 1000 MW(e). But, a latest and typical design (FUJI-U3) is 200 MW(e), which can be categorized
as small-sized reactors with modular designs (SMR). The thermal output of FUJI-U3 is 450 MW(t) and thus a
44% thermal efficiency can be attained. In addition, the simple core structure and high fuel efficiency should
facilitate a favourable economic performance.
Molten fuel salt can contain thorium (Th) as fertile material and 233U as fissile material, and the FUJI-U3

273

design can attain a self-sustaining fuel cycle with a conversion factor of 1.0. Since MSR-FUJI applies the Thcycle, generation of plutonium (Pu) and minor actinide (MA) is very small compared with Light Water
Reactors (LWR). Furthermore, it can consume Pu, and can thus contribute to reduce the proliferation risk
caused by Pu from LWR spent fuel. It can also be used to transmute long-lived MA to shorter ones.

2. Target Application
MSR-FUJI can be applied not only to electricity generation, but also to transmutation of Pu and/or MA. Besides
these purposes, it can be used as a heat source for water supply by desalination of seawater or for hydrogenproduction, utilizing its high exit temperature of 704°C.

3. Design Philosophy
The design philosophy of MSR-FUJI is to achieve a high level of safety, good economic performance,
contributing to non-proliferation, and to achieve fuel cycle flexibility.
MSR-FUJI is based on the ORNL’s results, and has been optimized as a small sized plant and further simplified
by removing the online reprocessing facility. Based on the operating experience at three experimental MSRs
in ORNL, it has been verified that MSR-FUJI is feasible. The steam generator (SG) is however a major
unverified component but it can be developed based on Fast Breeder Reactor (FBR) experience and the recent
supercritical power station technology.
MSR-FUJI adopts a passive safety system to improve the safety, reliability as well as the economics. Molten
fuel salt can be drained to a sub-critical drain tank through a freeze valve. Since gaseous fission products (FP)are always removed from molten fuel salt, the risk at accidents is minimized. MSR-FUJI is operated at very
low pressure (0.5 MPa), and a thick reactor vessel and pipes are not required. There are no fuel assemblies orcomplex core internal structure, with the only component of graphite moderator within a reactor vessel. Basedon these design principles, in-factory fabrication would be simple.

4. Main Design Features
Power Conversion
The nuclear steam supply system (NSSS) consists of a reactor core, pipes, pumps, a heat exchanger (HX), anda steam generator (SG), which supplies steam to a turbine/generator (T/G). The schematic diagram of MSRFUJI shows only one loop, but a loop can be redundant depending on a plant size or a need for flexibility.
MSR-FUJI is designed to produce an exit temperature of 704°C in molten fuel salt, and its heat is transferredto the secondary salt through a HX. Then, its heat produces 538°C supercritical steam of 252 kg/cm2 at a SG,
and generates electricity by a supercritical T/G. Owing to its high temperature, MSR-FUJI can achieve 44%
thermal efficiency.
The primary loop (molten fuel salt loop) is operated with forced circulation by a centrifugal pump during
normal operation. The system also has a natural circulation capability in emergency conditions.

Reactor Core
The core structure is made of hexagonal shaped graphite moderator blocks. The blocks contain holes that serveas the flow paths of the molten fuel salt that flow upwards through the blocks circulated by the primary pump.
The molten fuel salt then goes to a heat exchanger to transfer the heat to the secondary coolant salt.
The concentration of the fuel composition can be adjusted at any time through the fuel concentration
adjustment system. Since there are no fuel assemblies in the core, refueling shutdown is not required, andcontinuous operation is possible. In order to achieve a core conversion factor of 1.0, it is recommended torefresh the fuel salt every 7 years. Periodic maintenance shutdown will be required as in any power plant

Reactivity Control
Reactivity control for long-time operation can be performed anytime by a fuel concentration adjustment
system. In normal daily operation, reactivity or power level can be controlled by core flow or by coretemperature. Control rods are withdrawn in normal operation and are inserted by gravity in case of emergency
shutdown.

Reactor Vessel and Internals
The reactor vessel is made of Hastelloy N. Since the operating pressure is very low (0.5 MPa) a ‘pressure
vessel’ is not required. The only one core internals is the graphite moderator blocks.

Fuel Characteristics
The molten fuel salt is a liquid form of fluoride (LiF-BeF2) with ThF4 and a small amount of 233UF4. A typicalcomposition is LiF-BeF2-ThF4-233UF4 (71.76-16-12-0.24 mol%).
Molten fluoride can be used at very low pressure owing to its very high boiling temperature and very low
vapor pressure. The melting temperature of the above fuel composition is 499°C. It can dissolve uranium (U)or Pu as fissile material so that low enriched uranium (LEU) or Pu can be used. Fuel assembly fabrication is
not required for molten fuel salt, and radiation damage or fuel cladding failure does not occur.
274

5. Safety Features(a) Engineered Safety System Approach Configuration
In case of pipe break, leaked molten salt is drained to an emergency drain tank without passing through a freeze
valve. A pressurization accident is very unlikely owing to its low vapor pressure. Therefore, an emergency
core cooling system (ECCS), containment cooling system (CCS), makeup water pools, and automatic
depressurization system (ADS) are not required. In order to protect against a freeze accident in a molten fuelsalt loop, a high temperature containment is equipped with heaters.

(b) Decay Heat Removal System
In normal shutdown condition, decay heat is transferred to a secondary loop and a steam-line loop, anddisposed to the ultimate heat sink (seawater etc.). If all pumps in a primary or secondary loop stop, fuel salt isdrained to a drain tank through a freeze valve. Decay heat at the drain tank is cooled by a passive heat removal
system, and finally its heat is disposed to the outside environment through an air-cooled system that does not
require electricity.

(c) Emergency Core Cooling System (ECCS)
As is explained above, redundant and diverse ECCS and makeup water pools are not required. This would
simplify the plant, and eliminate concerns offailures in safety systems,(d) Containment System
Since the risk of pressurization accidents is very
unlikely, the containment size can be
minimized. Although molten salt is not
flammable, inert gas (N2) is enclosed within acontainment in order to maintain fuel salt purity
in case of a pipe break accident. The MSR-FUJI
design has 3 levels of containment. The 1st is
the reactor vessel and pipes made of Hastelloy
N. The 2nd is a high temperature containment
composed of three layers, which contains areactor vessel, pipes, and a heat exchanger. In
order to avoid a freeze accident, thiscontainment is equipped with heaters. The 3rdlevel is a reactor building composed of twolayers. As explained above, a pressurization
accident is very unlikely due the low vapor
pressure. Therefore, a containment coolingsystem and makeup water pools are notrequired.

Vertical cross-section of primary system of MSR-FUJI

6. Plant Safety and Operational Performance
Overall safety is described above. In case of a station blackout (SBO: Loss of all AC electricity) the MSRFUJI can be shut down and cooled without electricity. Core meltdown or steam/hydrogen explosion is
physically excluded by design, and no ECCS is needed. Long-time and daily operations are described in
Section-3(d). Based on those features, load following is performed without control rods.

7. Instrumentation and Control Systems
Instrumentation and control (I&C) systems in the MSR-FUJI design are the same as recent LWR designs. It
must support operators in making decisions and efficiently operating the plant during plant start-up, shutdown,normal operation, surveillance testing, and accidental situations. It adopts the man-machine interface more
useful, and expands the scope of automatic control.

8. Plant Layout Arrangement
Major buildings of MSR-FUJI are a reactor building, a SG building with a main control room, and a T/Gbuilding.

275

(a) Reactor Building
The reactor building contains a high temperature
containment, drain tanks, a radio-waste storage,
and other facilities required for the reactor. This
reactor building is a cylindrical shape with ahemispherical dome, which is made of concretewith steel liner as its inner layer. The reactor
building is founded on a common base-mat
together with other buildings.

(b) Control Building
The main control room (MCR) is located at a SG
building, which is next to a reactor building. The
MCR is a key facility to cope with normal andemergency situations, so it is designed to ensurethat plant personnel successfully perform thetasks according to the proper procedures.

Bird's-eye view of MSR-FUJI(c) Balance Plant
Turbine Generator (T/G) contains the supercritical T/G to produce electricity. Also, it contains condensers for
disposed steam, which use outside water (sea water etc.) for cooling. The electric power systems include themain generator, transformers, emergency diesel generators, and batteries, besides external power lines. In caseof station blackout, it can be shut down and cooled without electricity.

9. Testing Conducted for Design verification and Validation
MSR-FUJI is based on 3 experimental reactor experience. But some of technologies such as a supercritical SG
or remote maintenance equipment have to be established.

10. Design and Licensing Status
Preliminary designs for various applications have been completed [1]. Three experimental MSRs wereconstructed, and one of them was operated for 4 years without severe problems. Although the detailed design
is not yet started, safety criteria and guidelines for MSR licensing are proposed with numerical results for
major accident analysis [1].

11. Fuel Cycle Approach
As explained in Section-1 and 3, FUJI-U3 design can attain a self-sustaining fuel cycle. MSR-FUJI is
simplified by removing the online reprocessing facility. Only gaseous fission products (FP) are always
removed from molten fuel salt. Spent fuel salt is discharged, and reprocessed at the off-site pyro-reprocessing
facility. Some usable actinides can be sent back to the reactor.

12. Waste Management and Disposal Plan
Actinides such as U/Pu/Th/MA separated at the off-site reprocessing facility are recycled to MSR. FPs andsalt are stored at disposal facility. These facilities are to be developed in parallel with MSR deployment.

13. Development Milestones
1980’s1980’s
Until 2008
RecentConceptual designs of MSR-FUJI have been started
Accelerator Molten-Salt Breeder (AMSB) design for a large production of fissile material(similar to Accelerator Driven System ADS)
Designs such as a pilot plant (mini-FUJI), a large-sized plant (super-FUJI), a Pu-fuelledplant (FUJI-Pu).
The latest SMR plant (FUJI-U3).

[1] Yoshioka, R., Kinoshita, M. “Liquid Fuel, Thermal Neutron Spectrum Reactors”, Chapter-11 of the book “Molten Salt Reactor andThorium Energy”, Elsevier Inc., USA, 2017(1st edition), 2022(2nd edition).

276

THORIZON(THORIZON B.V., Netherlands)
All content included in this section has been provided by, and is reproduced with permission of THORIZON B.V., Netherlands.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
THORIZON, Netherlandsof origin
Reactor type
Coolant/moderator
Thermal/electrical capacity
Primary circulation
Primary system operatingpressure

Modular core molten salt reactor
Molten salt/graphite (no salt-graphite
contact)
100-300 max MWth 550 C steam/ 40120 max MWel
Forced circulation
2 bar max at shutdown – 8 bar max
under operation

Core inlet/outlet coolant
temperature (oC)

500 / 800

Fuel type/assembly array

Fissile-fertile material bearing molten
salt in closed containment modules

Number of fuelassemblies/modules in the core

3-7 (current designs, can change)

Fuel
Refuelling cycle
Reactivity control mechanism

Plutonium-Thorium
5 – 10 years target
Control rods, burnable poison, coredraining
andstrongnegativetemperature feedback

Approach to safety systems

Passive as much as possible

Design life (years)

full core replacement strategy allows
life extension beyond 60 years)
LWR-Plutonium burning, Thorium
breeding-burning
Nuclearsafe
corematerial
replacement strategy, with externalcore module series production,
continuous improvement by module
updates, and fuel cycle flexibility
Conceptual design

Fuel cycle
requirements/approachDistinguishing features

Design status

1. Introduction
Thorizon develops a novel molten salt reactor design, based on a modular core approach, in which the core
consists of multiple individually contained replaceable modules. When the modules are removed, the reactor
has no primary circuit and no primary circuit components. The reactor has no fixed large reactor pressure
vessel. The technology basis is flexible and allows larger and smaller systems, either by larger or smaller
modules, or by more or less modules. The technology offers fuel cycle flexibility, with first goal to burn LWRplutonium and commercialize, which can be followed by advanced versions on the same technology basis inwhich fuel cycles can be closed. Thorizon has finalised the conceptual design phase, has a positively evaluatedpatent, and successfully passed technical due diligences by third parties, and has secured the finances to enter
the detailed design phase and execute a core material irradiation program.

2. Target Application
THORIZON offers a technology base for a multitude of different molten salt reactor designs. First goal is tomaximize efficient LWR-Plutonium burning within the shortest route to deployment, to reduce the long-lived
nuclear waste burden, and turn waste cost into energy generation income. The technology basis combined withthe demonstration and commercialisation of Plutonium burners, could allow and support further expansion
towards a closed Thorium cycle, or fast spectrum options.

3. Design Philosophy
The largest challenge in molten salt reactors is related to the degradation of core materials. Thorizon’s design
tackles this challenge by allowing convenient and safe replacement of core materials, while maintaining277

containment. This approach has led to a design that offers additional advantages, in terms of flexibility, plant
lifetime, improvement implementation over the plant lifetime, cost, modularity/series production and practicalmanagement of spent fuel and degraded core materials. Another key aspect of the design approach is to avoid
complexity, eliminate proliferation risk concerns, and make sure that the design safety and performance canbe assessed by basic computational tools, facilitating validation and licensing.

4. Main Design Features(a) Power Conversion
Thorizon systems produce 550°C supercritical steam in a tertiary circuit, connected to a molten-salt based
secondary system by heat exchanger, connected to the primary loops of the modules, each with their own heatexchanger. The steam can be used directly in industrial processes or can be used to generate electricity in a
steam turbine. The secondary-tertiary system is largely based on existing technology from concentrated solar
power applications.

(b) Reactor Core
The reactor core consists of the upper sections of the modules, which when placed together form a criticalzone. Each individual module is subcritical, unless placed in a critical configuration with other modules andmoderator/reflector. For IP protection and export control reasons, Thorizon does not disclose information on
design geometry and materials.

(c) Fuel Characteristics
Thorizon aims to adopt LWR-plutonium and Thorium as a fissile-fertile fuel mixture. Information on exact
salt composition is regarded IP and export control sensitive.
Based on successful implementation and commercialisation, advanced modular core options can beconsidered, adopting different fuel cycles.

(d) Reactivity Control
Reactivity control is performed with burnable poison, control rods and passively via large negative temperature
feedback. The latter is a specific benefit of molten salt reactors, and utilised in the Thorizon design. To achievea long cycle, the initial over-reactivity is reduced by burnable absorber in replaceable core materials.
Reactivity is quickly reduced in case module pumps stop, as this leads to immediate core draining. Control rod
insertion, also renders the reactor subcritical.

(e) Reactor Pressure Vessel and Internals
Thorizon has no reactor pressure vessel, but a radiation shield which also serves as external impact barrier, inwhich modules are placed that each have their own two containments. Each module contains its own pumpand heat exchanger, and is closed.

(f) Reactor Coolant System
The primary circuit consists of individually contained modules, in which primary fissile-fertile material
bearing molten salt is circulated, for each module individually. The number of modules can be varied, as can
the size of the module, leading to different system powers. Each of the modules is connected to the same
secondary cooling system.

(g) Secondary System
The secondary system is a central molten salt system, which connects to each module where the salt is heated,and connects to the heat exchanger of the steam generator where it is cooled down. Optionally the secondary
system includes a heat storage system.

(h) Steam Generator
The secondary salt transfers the heat to a steam generator, providing 550°C steam, which can be used forindustrial process directly, or is used for high efficiency electricity production. The steam generator and energy
conversion system is off-the-shelf conventional.

5. Safety Features(a) Engineered Safety System Approach and Configuration
Molten salt reactor technology offers passive safety features, which are maximally exploited by the Thorizon
design. In addition, the low pressure in the primary system, which can be secured under all circumstances,
reduces the stresses on primary components, and excludes having a significant force to release materials to theoutside. Defence in depth is incorporated by adopting multiple containments.

(b) Decay Heat Removal System / Reactor Cooling Philosophy
Thorizon systems adopt an active secondary (molten salt)-tertiary (water/steam) coolant system, and a passive(inert gas) core cavity cooling system.
278

Decay heat in a station blackout scenario is passively removed from core modules by gas flow of the corecavity coolant system along the modules, which has adequate capacity avoiding temperature escalation in thecore modules. The system power, and component sizes, cooling surfaces and core cavity cooling system havebeen designed for that purpose. The core cavity cooling system also passively cools the outside of the modules
and the moderator and reflector in between and around the modules, during operational conditions.
Considering this is worst case, adequate cooling can always be maintained via this system, or via the normal
secondary-tertiary cooling system under normal or off-normal operating conditions.

(c) Spent Fuel Cooling Safety Approach / System
Temporary onsite storage of the core modules is assumed necessary to minimize core replacement outage
times. Core section materials are considered medium term waste. Spent fuel salt will be reprocessed after
extraction at the reprocessing plant. Non-core section materials can be considered for recycling, economics ofwhich need to be determined.

(d) Containment System
Defence in depth is managed by multiple barriers, taking credit for the absence of primary circuit pressure
escalation scenario’s under all circumstances:- Molten salt matrix (solid or liquid)- First and second containment for each core module- First and second containment for the secondary coolant circuit- Irradiation shield and external impact barrier surrounding the modular core
- Underpressurized building enclosure(e) Chemical Control
Chemical control is passive, no elements are introduced or extracted from modules during operation. Chemical
control methodology cannot be disclosed at this time, to protect IP.

6. Plant Safety and Operational PerformancesCore damage frequency has not been determined yet. It is however to be noted that pressure escalation in the
primary containments can be excluded, with a secondary containment to avoid release in case of primarycontainment failure.
Passive safety features, passive reactivity control and external impact barriers, largely reduce core damage andrelease probability to very low levels.
Station black-out scenario is managed by passive cooling as well, providing indefinite grace time.

7. Instrumentation and Control System
Plant detailed design is under development. Reactor control and shutdown systems have redundancy.
Instrumentation relates to temperatures and pressure in primary, secondary and tertiary circuits.
Activity monitoring of secondary and tertiary system, and the core cavity cooling system, and of the space inbetween primary module containments.
Flux measurement in the moderator-reflector.
Chemical potential of the primary salt is foreseen to be monitored as well, as measure of primary salt
corrosiveness and changes thereof.

8. Plant Layout Arrangement
Without modules present, the plant has no primary circuit or primary circuit components in the plant, and cantherefore ne considered largely conventional. Primary system components are manufactured in series andfuelled outside the plant.
At this stage, the purification and conditioning of primary salt is excluded, hence the plant has no salt
reprocessing equipment or facility on site, other than what is included inside modules.
Current design efforts focus on the primary system, core cavity cooling, reflector-moderator and radiation
shielding, to be expanded first to the secondary and tertiary cooling and heat transfer systems. Enveloping
building design and control room location and design is under development.

9. Testing Conducted for Design Verification and Validation
An irradiation program of core materials has been initiated. This program will provide material properties asfunction of irradiation damage and temperature, as a basis for safety and performance analyses, and to
determine expected module lifetime.
Primary system thermalhydraulic calculations are to be validated with scaled mock-up and a test loop (bothunder development).
The Thorizon concept allows for non-nuclear validation and qualification by a single module mock-up, being
representative for the whole reactor system, facilitating the validation, qualification and licensing process. Thisis foreseen to be established after the current final design phase is finalized.
The design has been analysed by thermalhydraulic system code and deterministic and probabilistic core
physics codes, reproduced by a qualified external party, confirming no safety showstoppers.
279

10. Design and Licensing Status
Interaction with Nuclear Safety Authority has started; Site selection is underway, not to be disclosed at thisstage. License to build is targeted for 9 years from now, license to operate 14 years from now. It is Thorizon’s
aim to accelerate this schedule by co-development with industry as soon as possible, to initiate component
manufacture, optimisation and qualification quickly.
The current phase is intended to be executed by a dedicated in house team of 15 people, supported by external
consultants for specific subjects, for which the funding has been secured. Thorizon aims to establish
partnerships and to raise additional funding, to support acceleration of the current schedule. Thorizon prefers
not to share conceptual technical information in much detail, due to IP protection and export control
considerations. The image shown on the first page is a high level, schematic visualisation, showing the module
concept and approach, without disclosing technical detail.

11. Fuel Cycle Approach
Modules are replaced as a whole, and salt maximum burn-up is matched by core material maximum lifetime.
A 5-10 year cycle is foreseen (with reactor at maximum power 100% of the time). Reactor criticality can matchthese cycle lengths with the fuel and core design foreseen, with material lifetime confirmation required bymaterial test irradiations.

12. Waste Management and Disposal Plan
By design, Thorizon systems do not release liquid or solid radioactive waste during normal operation. Afteruse and cooling, the modules can be transported in shielded casks of manageable dimensions, to be transported
to waste treatment facility, where the salt is extracted for separation and purification and recycling, the corematerials are cleaned and prepared for medium term storage. Economics and feasibility of ex-core sections(low-medium level waste) recycling has not yet been determined.

13. Development Milestones
2020

2022
2025
2030
2035
>2035

280

Conceptual design phase, technical due diligence successful, patent positively evaluated,investment secured
Transition from phase 1 conceptual design to phase 2 final design
Phase 2 final design finalised, all safety evaluations and PSA successful, license
application submitted, core material irradiation test program finalised, site selected, salt
experimental program initiated, start of phase 3
Finalisation of Phase 3. Salt experimental program finalised, salt manufacturing ready,transport casks ready. Components qualified, site prepared, module non-nuclear
prototype demonstration successful, license to build, Phase 4 construction starts.
Construction finished, license to operate
Initial operation experience, commercialisation

Stable Salt Reactor - Uranium (SSR-U)(Moltex Energy, United Kingdom)
All content included in this section has been provided by, and is reproduced with permission of, Moltex Energy, United Kingdom.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
Moltex Energy, United Kingdom
country of origin
Reactor type
Coolant/moderator

Molten Salt Reactor
Molten eutectic fluoride coolant salt / solidcommercial grade graphite moderator

Thermal/electrical
capacity, MW(t)/MW(e)Single unit – 40 / 16, deployable in arrays

Primary circulation
NSSS Operating
Pressure(primary/secondary),
Core Inlet/Outlet

Natural convection loop
Primary and secondary coolants atmospheric/ unpressurised (0.1 / 0.1)
725 / 795

Coolant
Temperature(oC)
Fuel type/assemblyarrayPower conversion
process

Molten salt fuel within vented fuel tubes
hexagonal
Moltensalt to arraysteam to turbine

Fuel enrichment (%)
Refuelling cycle(months)
Core Discharge Burnup(GWd/ton)

6
240
70

Reactivity control
mechanismApproach to safetysystems
Design life (years)
Plant footprint (m2)
RPV height/diameter(m)
RPV weight (metric ton)
Seismic design (SSE)

Strong fuel temperature coefficient, liquid
neutron absorber thermometer
Hazard reduction, inherent/passive safety
60
1 unit ~100, 32units (0.5 GW(e)) ~ 10,000-7.5 / 6 (low pressure)
Empty ~100 / Filled ~500
EUR standard - peak ground acceleration of
0.3g horizontally and vertically

Fuel cycle requirements/ approachOnce-through fuel cycle. All fuel replaced
together, 3 cycles total
Low pressure molten salt reactor system; ~20years fuel cycle; ~700°C heat output; Thermalenergystorage to allow operation as peaking
Basic design

Distinguishing features
Design status

1. Introduction
Moltex has the technology to transform theenergy landscape, ushering in a new age of clean,
low-cost energy for all. Moltex's Stable Salt
Reactor-Uranium (SSR-U) is an innovative
modular molten salt reactor, with each reactor
module providing 40 MW(th)/16 MW (e). Its
simple design means that most components canbe factory-produced and assembled on site,enabling a rapid roll-out of the technology at scale
and pace.
The SSR-U's unique attributes include; ~ 20 years
fuel cycle using Low Enriched Uranium (6%),
high temperature heat output (700oC) whichenables thermal energy storage in molten salt andoffers high quality process heat including thermochemical production of synthetic fuels.
The reactor is fuelled for a life of 16 full power

Schematic of an SSR-U Reactor single unit

281

years at commissioning, with a further 2 fuelling cycles possible, giving the SSR-U one of the longest
refuelling periods (comparable to HALEU reactors), and resulting in a total reactor lifetime of circa 60 years.
The simplicity of the design, low pressure operation, inherent and passive safety features, in combination withthe refuelling requirements result a low capital and low operating cost reactor.

2. Target Application
The reactor is flexible in its application, it is suitable for grid connection or standalone electricity and heat
supply. The high temperature output can be used to provide economical thermal energy storage in molten salt,
known as GridReserve®, which can then be despatched via normal steam and turbine generator arrangements
to meet fluctuations in demand, for example due to the intermittency of renewable energy supplies.
It can also be used to directly feed industrial processes, including thermo-chemical - hydrogen production,
which offers increased efficiency compared to electrolysis. The hydrogen itself can be used in several ways tosupport domestic, commercial, and industrial applications, such as; a direct substitute for gas use in industry,domestic heating fuel, for fuel cells for heavy transport, and as a feedstock for synthetic fuels, including
ammonia as a bunker-oil substitute for shipping. The reactor is also well suited to combined heat and power
generation for district heating.

3. Design Philosophy
Moltex' design philosophy is predicated by reducing cost, bringing the technology to market as quickly as
possible, and developing a reactor that can be deployed worldwide to meet its mission of having a meaningful
impact on global CO2 emissions. Cost reduction is achieved through hazard reduction and elimination, andthe use of inherent and passive features to deliver safety.
Moltex' approach is to adopt solutions that are already proven in the nuclear sector, meaning a significant
proportion of the design is already mature. The use of proven nuclear materials and well validated modelling
tools, mean that in combination, it is possible to rapidly move through the Technology Readiness Levels(TRL). Global deployability is ensured through the use of LEU, and its modular construction.

4. Main Design Features
Power Conversion
Each SSR-U reactor provides output of 40 MW(t) as hot salt at 700oC, which equates to ~ 16 MW (e). This
heat can be directly fed to a steam boiler to generate steam at typical subcritical conditions where there isalready a wealth of experience in flexible operation of generating plant. Alternatively, the heat can be fed toexternal thermal storage tanks (located outside the nuclear island), known as the 'GridReserve®' system.
The use of the GridReserve® technology permits electrical generation to follow a dispatched load profile orto operate in a frequency responsive mode. It is also possible to dynamically allocate energy to other heatloads. The ability of one or more SSR-U reactor systems to operate with different sizes of GridReserve® meansthat different energy load profiles are readily accommodated.

Reactor CoreThe SSR-U is a fluoride salt reactor with separate fuel and coolant salts. The core comprises an array of fueltubes in a graphite matrix which fills most of the tank. Each tube sits in a separate channel and molten salt
primary coolant circulates up through the channel by natural convection. It then flows through a heat exchanger
and flows around the core back to the fuel channels.The fuel salt is contained in vented fuel tubes. The fuelsalt circulates in the fuel tube by natural convection and heat is transferred through the tube wall to the coolant
surrounding the tube. The fuel salt does not mix with the coolant salt under normal operation. Under fault
conditions the miscibility of fuel and coolant salt has a fundamental impact on the concept of core damage. In
a conventional reactor, core damage can result in the reactor moving to a substantially more hazardous state.
In the SSR-U, similar scenarios are tolerable since even severe core damage moves the reactor to a safer state.
Breach of the tube wall, for instance caused by the tube melting, results in a reduction in core reactivity as thefuel salt is diluted in the large coolant volume. As a result, core damage does not result in an increase in theprobability of radiation exposure of workers or the public.

Reactivity Control
In the SSR-U, the bulk of the excess reactivity present immediately after fuelling is neutralised through fixedburnable absorber in inserts within the graphite matrix. Any residual excess reactivity is neutralised by pellets
of neutron absorber which are soluble in the coolant salt and added to the primary coolant using a periodic
injection system. This neutron absorber burns out slowly and is replenished at intervals of several weeks ormonths. Fine reactivity control for reactivity changes during power demand shifts is provided by a patented
novel mechanism akin to a conventional mercury thermometer, but instead filled with a liquid neutron
absorber. The “bulb” of the thermometer sits in the hot coolant salt exiting the top of the fuel tube channel
and the stem extends through the graphite matrix towards the centre of the core. An increase in heat demand
from the reactor causes a drop in primary coolant temperature and as a result the bulb of the thermometer cools,
causing the neutron absorber to be withdrawn from the stem of the thermometer and increasing reactivity andpower output. Conversely, when heat demand is reduced, the temperature of the primary coolant rises, heating282

the thermometer bulb, driving the neutron absorber through the stem, and thus, reducing reactivity and power

Reactor Vessel and Internals
The reactor vessel for the SSR-U comprises an outer and inner tank, separated by a refractory layer. The outertank constructed from a stainless steel will serve as the primary vessel for the reactor core and primary coolant
loop. The inner tank, also constructed from a stainless steel, will serve as the liner to contain the primarycoolant salt at temperatures above 700oC, and will be designed to accommodate radial expansion. The
refractory layer will act as the insulation layer for the inner tank, whilst also providing neutron irradiation
shielding to the outer tank. The inner and outer tank vessels are placed in a concrete pit underground andcovered from the top with a concrete shield.

Fuel Characteristics
The fuel is low-enriched uranium fluoride salt (6% enrichment), which is contained at low pressures by thepatented vented fuel tube.

Fission Product Management
Many of the fission gases are immediately captured as non-volatile salts and remain contained in the fuel salt
within the fuel tube. The major gas releases are the Xe and Kr noble gasses and the great majority of theradioactive species decay within the tube and bubbler.

5. Safety Features
Engineered Safety System Approach and Configuration
The design philosophy adopted is to follow theinternationally accepted principle of the risk
mitigation pyramid. The focus is to eliminate hazards
wherever possible, to use inherent and passive safety
features where this is not possible, and only to rely on
engineering or administrative controls when thatcannot be achieved.

Decay Heat Removal System
The decay heat removal system is completely passive
and utilises natural convection with no need formechanical components. Utilising an arrangement of
perforated ducts, it promotes air cooling of the baseand side walls of the reactor tank, drawing cool air
down and around the external tank walls andexhausting the heated air to the reactor building.

Internationally recognised Hierachy of Controls

Emergency Core Cooling System
A traditional Emergency Core cooling system is not be required for the SSR-U as the design ensures that thecore is never exposed by the coolant, and decay heat is removed by convection as previously explained.

Containment System
Containment is progressivley provided by the fuel tube, the coolant salt, the inner tank, the refractory insulationof the tank, the outer tank, and the concrete bilogical shield.

Chemical Control
The SSR-U has distinct fuel and coolant chemical environments. The fuel salt chemistry is very simple and isredox stabilised by a mixture of uranium oxidation states acting as a buffer in a eutectic mixture with metalfluoride diluent. The redox buffer enables maintenance of redox potential and mitigation of potentiallycorrosive fission products generated through life. Trials to demonstrate suitability of fuel chemistry are
underway. The primary and secondary coolant are both fluoride eutectics. Tritium production is mitigated byeliminating Lithium from the chemistry. Similarly, Beryllium is also eliminated to prevent highly corrosive
and toxic environments. Both coolant salts are compatible with graphite and mix safely with the fuel salt.
Corrosion of the reactor components by the coolant salts is managed using reducing agents. Some corrosion
testing of candidate fuel mixtures has already been carried out with plant materials at representative reactor
temperatures.

Spent Fuel Cooling Safety Approach / System
Once the SSR-U's fuel is depleted, a cooldown operation occurs before the fuel is removed from the reactor.
During cooldown, there is no requirement for additional cooling systems. The normal heat transfer systems
are disengaged and the decay heat removal system, as described in (b), provides sufficient heat removal until
the fuel is ready for extraction.

283

6. Plant Safety and Operational Performances
The design philosophy is such that there will be very limited operational requirements to attend the nuclear
island during each fuel cycle. The significant reduction in the quantity of engineered safety and component
systems will also substantially reduce the number of operating staff required for maintenance.
The SSR-U can ramp up and down in output, but Moltex considers the use of GridReserve® a better option
for the provision of flexible electricity generation, with the turbine hall providing more responsive ramp rates.
The SSR-U reactor system can be designed to operate at peak output throughout the ~ 20 year fuel cycle.

7. Instrumentation and Control System
Instrumentation is limited to flux monitors for controlled reactor commissioning, and temperature sensing toinitiate the periodic injection of pellets of neutron absorber for reactivity management as detailed in (c).

8. Plant Layout Arrangement
The SSR-U is flexible and can be deployed indifferent array sizes as demanded by theapplication. An illustrative SSR-U plant layout
arrangement would be an array of 32 reactors on
the nuclear island, which will provide circa 1.2
GW(th) energy which can be converted into ~512MW(e) of electrical output. The external
GridReserve® storage tanks, turbine hall,electrical sub-station, and potential hydrogenproduction plant are located outside the nuclear
island. This configuration will enable a truly‘hybrid’ energy generation, storage, and release Artists impression of an array of 32 modules configured in line withturbine hall, GridReserve® storage facility and sub-station
scheme.

9. Testing Conducted for Design Verification and Validation
Current testing is focused on materials & corrosion, confirmation of thermo-physical properties of salts for use
in modelling, and thermohydraulic tests. Moltex Energy has established its own laboratory and is running a
range of corrosion and materials compatibility tests, alongside thermophysical properties tests. Industrial andacademic partners are supporting Moltex Energy with analysis of test samples including ICP-OES (for salt
quality) and SEM (for materials performance).

10. Design and Licensing Status
Interaction with Nuclear Safety Authority to be launched in 2023. Site permit for FOAK to be developed,
however a number of sites are under consideration and discussions are ongoing.

11. Fuel Cycle Approach
The SSR-U is fuelled for 16 full power years, with a further 2 fuel cycles possible. Interim storage of spent
fuel will be in multi-purpose storage cannisters on site until site decommissioning. The spent fuel will then be
conditioned for geological disposal or alternatively, recycled using another advanced nuclear technology underdevelopment by Moltex Energy, named WATSS (WAste To Stable Salt process).

12. Waste Management and Disposal Plan
Moltex are currently developing the design to be consistent with international good practice for waste
management and spent fuel, aligned to IAEA. The reactor system itself is being developed to design out asmuch operational, refuelling, and decommissioning waste as is reasonably practicable. The waste hierarchy issuperimposed across the design process, by first seeking to avoid waste creation, and then minimising thecategory of unavoidable waste. Moltex envisage minimal waste production during normal operation of the
SSR-U. Argon is likely to be generated as a by-product of the process, however, it is unlikely to ever exceed
the Argon discharge limits. Refuelling will generate Argon, Krypton & other gaseous wastes, but these will
not exceed their designated discharge limits. A small amount of solid waste will be produced through spent
fuel (which can be stored in multi-purpose cannisters on site until decommissioning or recycled using the
Waste To Stable Salt process -WATSS). Decommissioning will produce solid, liquid, and gaseous wastes, thelargest being graphite. Moltex seek to recycle and re-use graphite wherever possible in new SSR-U reactors.

13. Development Milestones
Q4 - 2022
Q3 - 2023
Q4 - 2023

284

Concept Design Freeze for SSR-U
Initiate regulatory asessment of the technology
Preliminary Safety Report (PSR) Complete

KP-FHR (Kairos Power, United States ofAmerica)
All content included in this section has been provided by, and is reproduced with permission of, Kairos Power, LLC, USA.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
Kairos Power, LLC, United
country of origin
States of America
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)Artist's rendering of the Hermes demonstration
reactor facility to be constructed in Oak Ridge, TN

Modular, pebble bed, hightemperature, salt-cooled reactor
Li2BeF4 (Flibe) / graphite
320 / 140

Primary circulation
NSSS Operating Pressure(primary/secondary) (MPa)

Forced Circulation< 0.2

Core Inlet/Outlet Coolant
Temperature (oC)

550 / 650

Fuel type/assembly array

TRISO particles in graphite
pebble matrix / pebble bed

Power conversion process
Fuel enrichment (%)
Refuelling cycle (months)
Reactivity control mechanism
Approach to safety systems
Design life (years)
RPV height/diameter (m)Seismic design (SSE)
Fuel cycle requirements /approachSuperheated steam cycle
19.75
Online refuelingControl elements, boron
Passive
20 (vessel), 80 (plant)
7.2 / 3.9
Target: contiguous USA
Once-through Uranium

Distinguishing features

Longer than 72-hour coping time
for core cooling without AC or
DC power, or operator action
Conceptual design

Design status

1. Introduction
Kairos Power is a mission-driven company singularly focused on its effort to commercialize advanced reactor
technology in time to play a significant role in the fight against climate change. Kairos Power is disrupting theindustry with rapid iterative development and vertical integration strategies to deliver a clean energy solution
with robust safety at an affordable cost.
The Kairos Power fluoride salt-cooled high temperature reactor (KP-FHR) is a novel advanced reactor
technology that leverages TRISO fuel in pebble form combined with a low-pressure fluoride salt coolant. The
technology uses an efficient and flexible steam cycle to convert heat from fission into electricity.

2. Target Application
This advanced reactor technology is designed for high availability and performance with low maintenance andlifecycle costs, providing dispatchable power that improves grid resiliency and security. In combination withvariable renewables, this technology can create a path to a truly clean energy system. The KP-FHR aims to becost competitive with natural gas.

3. Design Philosophy
The fundamental design concept is the combination of Tristructural Isotropic (TRISO) particle fuel coupled with molten
fluoride salt coolant (2LiF:BeF2, Flibe). This combination
results in a high temperature, low-pressure reactor with robust,passive safety systems. In addition to robust, inherent safety,
the design also reduces reliance on high-cost, nuclear grade
components and structures and leverages conventional
technologies to lower capital costs.

KP-FHR Power conversion system diagram285

4. Main Design Features
Power Conversion
The power conversion system will leverage prior technology demonstration of solar nitrate salt steam
generators and conventional power conversion technologies for the balance of plant.

Reactor Core
4.0-cm diameter spherical fuel and moderator pebbles forming the active region of the core. The core design
utilizes a cylindrical geometry with a graphite side-reflector and bottom and top graphite structures. The coreinternal structures enable reactivity control and shutdown elements.

Reactivity Control
Reactivity control for reactor maneuvering and non-accident events is provided by control elements that insert
into the graphite reflector surrounding the pebble bed core. Reactivity control during accident events isprovided by shutdown elements that insert directly into the pebble bed. The shutdown elements are gravity
driven and are released by the reactor protection system. Both the shutdown and control elements consist of acomposite structure of neutron absorber material made of natural B4C in an inert gas with SS316H cladding.
The shutdown and control elements fail safe (insert) on a loss of power. The number and the location of the
shutdown elements are selected to provide sufficient shutdown margin at zero-power conditions. The shutdown
system is designed with sufficient reactivity worth to shut down the core from hot full power assuming the
failure of the highest worth control element. The reactivity control elements can be inserted for shutdown at a
rate sufficient to assure that the design limits for the fission product barriers are not exceeded.

Reactor Vessel and Internals
The KP-FHR reactor vessel operates at low pressure, so has thin-walled construction using 316H stainlesssteel that complies with ASME Section III, Division 5 requirements. The reactor internals include a core barrel
and reflector support structure that position and retain the graphite reflector structure which maintainsalignment of the structure to form the major flow paths and reactor core volume. As the pebbles are positively
buoyant in the coolant, defueling occurs from the top of the core and online fuel addition from the bottom. The
reactivity control and other components interface through the top head.

Fuel Characteristics
Kairos Power’s reactor uses fully ceramic fuel, which maintains structural integrity even at extremely hightemperatures and will be undamaged to well above melting temperatures of conventional metallic reactor fuels.

Fission Product Management
The fundamental safety strategy for the KP-FHR is rooted in the retention of fission products within the TRISO
layers of the fuel particle design with additional retention in the flibe salt coolant. The TRISO layers are
credited for providing a ‘functional containment’ for meeting design basis accident dose limits. The siliconcarbide (SiC) coating on the TRISO particles is the primary fission product barrier, while the pyrolytic carbon
layers and matrix act as secondary barriers for trapping or impeding the transport of fission products andprotecting the integrity of the SiC layer.

5. Safety Features
Engineered Safety System Approach and Configuration
The reactor system includes a pebble bed core, surrounded by a graphite reflector, contained within acylindrical 316H stainless steel reactor vessel.

Decay Heat Removal System
Decay heat removal during normal operations and non-accident events is provided by a normal shutdown
cooling system that connects directly to the primary heat transport system. Decay heat removal during accident
events is provided by a passive reactor vessel auxiliary cooling system (RVACS) located external to the reactor
vessel. No coolant injection is required for inventory makeup nor decay heat removal functions under accident
conditions. The RVACS relies on decay heat through thermal radiation and natural convective heat transfer
utilizing a thermosyphon concept. The system evaporates water to provide heat removal and does not require
inventory makeup for a minimum of 72 hours. System valves are fail-safe and do not rely on safety related
electrical power or operator action.

Emergency Core Cooling System
Kairos Power reactors have uniquely large safety margins based on the selected combination of fuel andcoolant, which allows emergency cooling to be driven by fundamental physics rather than engineered systems.

Containment System
Functional containment in the KP-FHR is provided by the robust intrinsic safety characteristics of the TRISO

286

fuel and Flibe coolant to ensure that the health and safety of the public and workers are protected. Multiple
additional barriers in the KP-FHR provide defense-in-depth.

Chemical Control System
A Chemistry Control System (CCS) monitors the salt coolant to detect impurities, ensuring it stays within an
established operating window.

Spent Fuel Cooling Safety Approach / System
The zirconium cladding used in light-water reactor fuel contains substantial stored energy, which affects thesafety significance of spent fuel cooling systems. KP-FHR fuel is fully ceramic, which simplifies coolingrequirements for fuel in the pebble handling system and also in canister storage.

6. Plant Safety and Operational Performances
The KP-FHR leverages intrinsic safety characteristics of the fuel and coolant to achieve uniquely large safetymargins while lowering capital costs and improving operating economics. The fuel in the KP-FHR is the
TRISO particle fuel, which can withstand fuel particle temperatures up to 1600°C. The Flibe is chemicallystable and is at low-pressure, with a boiling point of 1430°C, notably lower than 1600°C and yet functionallyvery high. The combination of extremely high-temperature-tolerant fuel and low-pressure, single-phase,chemically stable reactor coolant removes entire classes of potential fuel-damage scenarios, greatly
simplifying the design and reducing the number of safety systems. The intrinsic low pressure of the reactor
and associated piping, along with the functional containment provided by the TRISO fuel, enhances safety andeliminates the need for high-pressure containment structures.

7. Instrumentation and Control System
The Kairos Power Instrumentation and Control Systems takes advantage of the inherent safety features of the
KP-FHR technology to simplify the reactor protection system configuration and automated actions. The system
is sufficiently simplified to allow for deterministic analysis of all design basis events. The plant control system
operates within the protection system established operating envelope as an industrial control system,
maximizing automation and continuous health monitoring of the plant. The I&C Reactor Protection System(RPS) provides protection during steady state and transient power operation and includes the capability to
manually or automatically trip the reactor and activate RVACS. The RPS is fully independent of the Plant
Control System, which provides overall control during normal operation, startup or shutdown.

8. Testing Conducted for Design Verification and Validation
Kairos Power is committed to a culture that embraces iterative development and facilitates an integrated design
philosophy, testing program, and licensing approach to mitigate technical, licensing, manufacturing, andconstruction risk while establishing cost certainty through iterative hardware demonstrations.
By mirroring SpaceX and its successful rapid spiral development, Kairos Power accelerates test cycles forinnovation and optimization. Our rapid iterative approach leverages multiple design-build-test cycles with both
nuclear and non-nuclear systems prior to the first commercial reactor. Major hardware iterations include:
Engineering Test Unit (ETU) – a
non-nuclear system that will test
mechanical designs, scale up Flibe
production, establish manufacturing
infrastructure, and build a sustainablesupply chain for raw materials and offthe-shelf components. ETU is set to becommissioned in Summer 2022.
Hermes – a 35 MWth non-powerreactor that will demonstrate Kairos
Power's ability to produce affordable
Kairos Power's Rapid iterative approach to hardware development
nuclear heat. Currently in the designphase, Hermes is scheduled to be operational in 2026.
U-Facility – a non-nuclear demonstration intended to test manufacturing and construction of primary reactor
systems, serve as a training center and reduce O&M cost uncertainty
KP-X – a first-of-a-kind 140 MWe commercial KP-FHR operating at grid-scale

9. Design and Licensing Status
Kairos Power submitted the construction permit application (CPA) for the Hermes low-power demonstration
reactor to the U.S. Nuclear Regulatory Commission (NRC) in October 2021 and it was accepted for review
the following month. Prior to that, we conducted extensive pre-application engagement with the NRC. To date,
Kairos Power has submitted 11 topical reports and several technical reports, many of which have been
approved by NRC staff with the remainder expected soon. The Construction Permit Application for the Hermes
low-power demonstration reactor is under review by the U.S. Nuclear Regulatory Commission. Scheduled to287

achieve criticality in 2026, Hermes is a reduced scale (35MWth) demonstration reactor that will prove Kairos
Power's capability to deliver low-cost nuclear heat.

10. Fuel Cycle Approach
The KP-FHR operates on a continuous refueling cycle. Fuel pebbles are extracted from the core in the reactor
vessel and inspected for burnup and integrity in a pebble handling system. Pebbles are either inserted back into
the active core or directed to spent fuel storage. Fuel pebbles pass through the core six times on average before
reaching full burnup.

11. Waste Management and Disposal Plan
The used TRISO pebble fuel and a small inventory of ‘greater than Class C’ waste is packaged in multi-purpose
canisters for dry interim storage and subsequent off-site transportation for direct geologic disposal or recycling
depending upon national policies. All remaining waste streams from operation and decommissioning qualify
for low-level waste disposal.

Engineering Test Unit under construction

Salt Lab located in Alameda, CA

RAPID Lab located in Alameda, CA

In-house manufacturing shop located in KP-Southwest

12. Development Milestones
2018
2018
2018
2020
2020
2020
2020
2021
2021
2021
2021
2022
2022
2022
2022
2022

288

Pre-Conceptual Design Completed
Commissioned R-Lab
Initiation of Pre-Application Review with NRC
Acquisition of KP-Southwest Facility in Albuquerque, NM
Commissioned S-Lab
Formed strategic partnership with Materion Corporation to produce Flibe
Awarded $629M cost-shared Advanced Reactor Demonstration Program Risk Reduction Award from the U.S. DOE
Cooperative Development Agreement with Tennessee Valley Authority to deploy Hermes low-power demonstration
reactor
Testing Facility construction completed in Albuquerque, NM
Hermes Construction Permit Application accepted for review by U.S. Nuclear Regulatory Commission
Construction begins on the Engineering Test Unit (ETU)
Kairos Power establishes the KP-OMADA Advanced Nuclear Alliance with leading North American utilities
and generation companies
30,000th pebble manufactured for ETU
Manufacturing infrastructure expanded at KP-Southwest[Planned Milestone] Molten Salt Purification Plant (MSPP)[Planned Milestone] ETU commissioningMk1 PB-FHR (UC Berkeley, United Statesof America)
All content included in this section has been provided by, and is reproduced with permission of, UC Berkeley, United States of America.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type
Coolant/moderator

University of California, Berkeley,
United States of America
Fluoride-salt-cooled hightemperature reactor (FHR)
Li2BeF4 / graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

236 / 100NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/ assembly array

0.3 / 1.85 (compressor outlet)

Fuel enrichment (%)

Forced circulation

600 / 700
TRISO particles in graphite pebble
matrix / pebble bed / 470 000
19.9

Fuel burnup (GWd/ton)
Fuel cycle (months)

180

Fuel core residence time(months)
Main reactivity control
mechanismApproach to safety systems

2.1, average of 8 passes to achieve
full burn up
Negative temperature coefficient;control rod insertion
Passive

Design life (years)

60

Plant footprint (m2)

45 000RPV height/diameter (m)12 / 3.5

On-line refuelling

Seismic design (SSE)~ 0.3gDistinguishing features

Large fuel and coolant thermal
margin, high temperature operation
Pre-conceptual design

Design status

1. Introduction
The Mark 1 Pebble-Bed Fluoride-Salt-Cooled High-Temperature-Reactor (Mk1 PB-FHR) is a small, modular
graphite-moderated reactor. FHRs are differentiated from other reactor technologies because they use hightemperature, coated particle fuels, and are cooled by the fluoride salt flibe (7Li2BeF4). The Mk1 PB-FHR design
described here is the first FHR design to propose driving a nuclear air-Brayton combined cycle (NACC) forbase-load electricity generation.

2. Target Application
The Mk1 PB-FHR is designed to produce 100 MW(e) of base-load electricity when operated with only nuclear
heat, and to increase this power output to 242 MW(e) using gas co-firing for peak electricity generation. This
provides a new value proposition for nuclear power to earn additional revenues by providing flexible gridsupport services to handle the ever-increasing demand for dispatchable peak power. This is in addition to
traditional base-load electrical power generation.

3. Design Philosophy
The Mk1 PB-FHR is designed with advanced passive safety features and intrinsic fuel and coolant properties
which make the consequences of severe accidents studied for light water reactors much easier to manage.
Passive safety mechanisms include natural circulation decay heat removal activated by a passive check valvein accident conditions and buoyant control rods for emergency shutdown without operator intervention.
Fluoride salt coolants have uniquely high volumetric heat capacity, low chemical reactivity with air and water,
very low volatility at high temperature, effective natural circulation heat transfer, and high retention of most289

fission products. These characteristics are in addition to reasonably low neutron capture probability (whenusing enriched 7Li), and good neutron moderation capability.

4. Main Design Features
Power Conversion System
The 236-MW(t) Mk1 PB-FHR uses the
NACC power conversion system. It uses a
General Electric (GE) 7FB gas turbine(GT), modified to introduce external
heating and one stage of reheat, in a
combined-cycle configuration to produce
100 MW(e) under base-load operation, andwith natural-gas co-firing to rapidly boost
the net power output to 242 MW(e) to
provide peaking power. The Power
Conversion Unit consists of the reactor
core, two coiled tube air heaters (CTAHs)
to transfer heat from the main salt topressurized air, a heat recovery steam
generator
system(HRSG),
steam
condenser, and the GT. During normal
operation, the primary coolant relies on
forced circulation

Reactor and Core-unit
The core design incorporates an annular pebble-bed core geometry composed of a homogeneous mix of fuelpebbles adjacent to the center graphite reflector, with a layer of inert graphite reflector pebbles on the outsidethat reduces the fast-neutron fluence to the outer fixed radial graphite reflector sufficiently for it to last the lifeof the plant.
The center reflector provides 8 channels for insertion of buoyant control rods, and it also provides flow
channels for radial injection of coolant into the pebble core, to provide a combined radial and axial flow
distribution that increases the effectiveness of heat transfer from the fuel and results in lower average fueltemperature. The center graphite reflector internals need to be replaced periodically due to radiation damage.

Reactivity Control
For reactivity control, the Mk1 is designed to have negative fuel, moderator, and coolant temperature reactivity
feedbacks. The design uses a buoyant control rod system for normal reactivity control, and the system also
provides a passive shutdown capability because the buoyant rods will insert if the reactor coolant temperaturein the control-rod channel exceeds 615°C, the buoyant stability limit. The design also uses shutdown bladesthat can insert directly into the pebble bed for reserve shut down. In the event that electrical power is interrupted
to the drive mechanisms for the motors of the control rod and shutdown blade cable drums, they will insert
and shut the reactor down.

Reactor Vessel and Internals
To enable near-term deployment, the Mk1 design uses a core barrel and other core internal structures fabricatedfrom the same metallic material as the reactor vessel and main salt piping. The outer radial reflector blocks arealigned and held against the metallic core barrel using a system of axial alignment ribs and radial retaining
rings quite similar to designs originally developed for the Molten Salt Breeder Reactor (MSBR) project. The
use of metallic core internal structures, rather than advanced ceramic composites, simplifies fabrication andlicensing for the Mk1 design.

Fission Product Management
The coated uranium particles are packed in an annular fuel zone around a low-density graphite core. One Mk1
pebble contains 1.5 g of uranium enriched in 235U to 19.9% and encapsulated inside 4730 coated particles. Thevery low circulating power for the coolant in salt-cooled reactors, compared to helium-cooled reactors, makes
it practical to use smaller pebbles. This small-pebble design doubles the pebble surface area per unit volume
and halves the thermal diffusion length, enabling a substantial increase in power density while maintaining
relatively low peak fuel particle temperature. The coated particles have the main fission retention function, but
the molten salt coolant, primary circuit and building also serve as barriers to release.

Heat Recovery Steam Generator System
The Heat Recovery Steam Generator System (HRSG) and steam condenser need to be sized for full power
operation at co-firing conditions. The large HRSG inlet temperature variation between baseload and co-fired
operation modes introduces certain caveats to the steam cycle design. With the expected frequent power
ramping of the GT and dissimilar ramping rates compared to the steam turbines/HRSG, special design290considerations are needed such as opening steam turbine inlet valves and or allowing some bypass flows.

5. Safety Features
The safety objective with the IMSR® design is to achieve high inherent safety, and a walk-away safe nuclear
power plant. No operator action, electricity, or externally-powered mechanical components are needed toassure the primary safety functions of controlling, cooling, and containing.

Engineered Safety System Approach and Configuration
For reactivity control, the Mk1 has a combination of intrinsic features and passive systems. It has negative
fuel, moderator, and coolant temperature reactivity feedbacks. The reduced fuel temperature in the PB-FHR
provides improved response to hypothetical ATWS accidents. The negative fuel temperature reactivity
feedback in FHRs is significantly larger than the coolant temperature reactivity feedback, because the coolant
does not boil—the boiling temperature of flibe is 1430°C—as in light water reactors (LWRs), and larger than
the graphite moderator temperature reactivity feedback. Under the beyond design basis ATWS accident wherereactor scram does not occur upon loss of flow or loss of heat sink, the FHR coolant equilibrates to atemperature close to the original fuel temperature. Simplified analysis for the Mk1 design indicates that thisequilibrium ATWS temperature will be below 800°C.

Decay Heat Removal System
In the PB-FHR core, the emergency heat removal safety function is also controlled by passive mechanisms.
The PB-FHR design concept employs a passive check valve to activate natural-circulation-driven heattransport to a set of three Direct Reactor Auxiliary Cooling System (DRACS) loops and ultimately to
Thermosyphon-cooled Heat Exchangers (TCHXs) upon Loss of Flow Condition (LOFC). Heat removal fromthe TCHXs is regulated by fail-open valves that supply water to the thermosyphons integrated into these heatexchangers. The valves are held closed during normal operation, and can also be closed to control over-cooling
during prolonged reactor shutdown. In addition to the passive emergency decay heat removal provided by the
DRACS, the PB-FHR power conversion system and the normal shutdown cooling system provide heat removal
capability and defence in depth in assuring adequate core heat removal.

Emergency Core Cooling System
Coolant inventory control is provided by fully passive mechanisms that require no RPS or manual operator
actions. The primary salt fulfils dual roles during design basis events, by providing natural-circulation heatremoval and preventing chemical attack to fuel pebbles from exposure to air. The PB-FHR utilizes a pooltype reactor configuration, similar to the design adapted for many sodium fast reactors. For BDBEs where thevessel leaks or ruptures, the Mk1 refractory cavity liner insulation system controls the level change in thevessel and prevents uncovering of fuel.

Containment System
The Mk1 design introduces another novel feature, a “gas gap” system, to make it physically impossible totransmit excessive pressures to the reactor vessel and reactor cavity/containment from potential tube or
manifold pipe ruptures in a CTAH. The gas gap is created adjacent to the containment penetrations for the hot
and cold legs. For the Mk1 PB-FHR, water pools are used inside the shield building to provide water tothermosyphon-cooled heat exchangers (TCHXs) in the DRACS modules, as well as to the reactor cavity liner
cooling system. Because these water pools also provide a source of water for evaporative cooling underbeyond-design-basis event (BDBE) conditions, they are provided with a secondary confinement following the“tank-within-tank” design principle.

6. Plant Safety and Operational Performances
Due to the high thermal efficiency of the NACC system, the steam-bottoming condenser requires only 40% of
the cooling water supply that is required for a conventional LWR, for each MWh of base-load generation. As
with conventional natural-gas combined cycle (NGCC) plants, this makes the efficiency penalty of using dry
cooling with air-cooled condensers much smaller, enabling economic operation in regions where water isscarce. The advantage of the NACC system arises from additional revenues earned by providing flexible gridsupport services because under base-load operation NACC power conversion has lower fuel costs than NGCC,
and under peaking operation has higher efficiency in converting natural gas to electricity than NGCC, NACC
plants will always dispatch before conventional NGCC plants.

7. Instrumentation and Control Systems
The digital control system is designed so that neither its actions nor its failure to act would have any deleterious
impact on the ability of the PB-FHR to respond safely to design basis events. The quality requirements for thecontrol system then arise from the economic incentives to maximize system performance and to preserve theinvested capital, thus high-quality commercial-grade equipment is anticipated to be used.
Except during startup and low-power conditions, the PB-FHR operates with constant core inlet and outlet
temperatures. Load-following capability is made possible by air bypass flow to respond to rapid load-changetransients and turbine inlet temperature control (by bypassing air around the CTAHs) for slower transients.
291

Pump speed control is then used to control the core temperature difference, and control rod position is used tocontrol the average temperature. The control system adjusts the pebble loading and unloading schedule tomaintain sufficient excess reactivity to accommodate a xenon transient equivalent to a rapid power reduction
from 100% to 40%.

8. Plant Layout Arrangement
The figure on the right presents a notional
180-acre site arrangement for a 12-unit
Mk1 PB-FHR power plant capable ofproducing 1200 MW(e) base load and2900 MW(e) peak power output. Due tothe much smaller cooling requirements,
they do not need to be sited near bodies ofwater. Population centers tend to belocated near bodies of water which meansthat FHRs can be sited in areas where
fewer people want to live. So, rather thanattempt to minimize the site footprint, themore important goal is likely to facilitate
construction of modules adjacent tooperating modules, and to optimize thedegree to which some services are shared.

Reactor Building

Construction
Area

Parking Lot andStorage (600 Spaces)

Primary Natural
Gas Isolation
StationSwitchyard

OutageSupportBuilding

North

Training
STG B

STG A
GSU

Shipping
andReceiving
Warehouse

GSU

Module Haul RouteConstruction
Parking
300 Spaces#8

CommonUnderground
TunnelConstruction
Material
Temporary
Storage

#7

#6

#5

#4

#3

#2

Mk1 PB-FHR Modules 75 m O/C
Spent Fuel Storage Pad
Space for 720 Dry Casks(90 depicted)
Rad
Waste
Bldg
#12

#11

#10

#9

Control
Building
Water Storage#1

Operator Parking
100 Spaces

FuelHandlingandStorage

Covered
Vehicle Inspection
StationParking

Protected
Area Entry
Bldg

Visitor Parking
100 Spaces

BackupGen
Hot/cold
MachineShops

Protected Area Fence
600 m x 200 m
Hydrogen
Storage

GSU

Salt
Storage

Main
Admin

STG C

Discharge Pond
Cooling Tower Group B

Scale (100 m per segment)

Underground common
utilities tunnel

Admin Parking
300 Spaces

Security
Checkpoint

Admin
Expansion

Water Treatment

Cooling Tower Group C

Cooling Tower Group A

Owner Controlled Area and Vehicle Barrier
950 m x 750 m

Intake ﬁlter
Main stack

The Mk1 reactor building and NACC
system arrangements supports a multi- Shield building
Simple cycle
module plant configuration by allowing DRACS chimney
bypass stackmultiple units to be lined up in a row with
Modiﬁed GE 7FB
a clear boundary between the reactor and Personnel airlockgas turbine
its vital areas, versus the balance of plant
Equipment
hatch
HRSG
(BOP). The GT and associated equipment
are configured to minimize the air pressure Fuel canister
Below-grade
loss and circulating power in the air well
air duct vault
ducting while maintaining a clear
Ventilation exhaustboundary between the reactors and the Gradesystem
BOP. This configuration makes it easier to levelco-locate combined nuclear services on
one side of a multi-module plant (training, fresh fuel handling/receipt, spent fuel dry storage, security, access
control, multi-module control room, hot-rad/Be shops, etc.), and have BOP combined services on the otherside (off-site transmission, process steam loads and/or steam bottoming turbines, cooling towers, etc.).
The Mk1 reactor building is partially embedded below grade, with the reactor deck located slightly above
grade, shortening the air duct lengths and the depth of the air-duct vault. The baseline Mk1 reactor building
design uses a cylindrical shield building fabricated from steel-plate/concrete composite (SC) modules, quite
similar to the Westinghouse AP1000 shield building. The overall height and diameter of the Mk1 shieldbuilding are 47.5 m and 24.5 m, respectively, compared to 83 m and 42 m for the 1150-MW(e) AP1000, so
the Mk1 shield building volume is 2.2 times greater than the AP1000, per MW(e) baseload.

9. Design and Licensing Status
Further work is needed in the definition and design of: plant staff capabilities and size, instrumentation
requirements, systems and equipment for operations and maintenance, future plant reliability and availability,
and licensing strategies for licensing commercial prototypes in the U.S. as well as internationally.

10. Fuel Cycle Approach
The initial design is based on once through LEU cycle but other thermal spectrum based fuel cycles (U-Th;
Pu-Th, Pu burner) as illustrated by HTR coated particle fuel, are in principle possible.

11. Waste Management and Disposal Plan
The fuel pebbles are based on HTR coated particle fuel with its excellent radioactivity containment
characteristics. Upon removal from the primary coolant, the spent fuel will require cleaning to remove residual
salt and cooling to maintain acceptable fuel temperatures in the gas environment.

12. Development Milestones
2018

292

Pebble bed FHR technology, with significant similarities to the Mk1 PB-FHR, is being
developed by Kairos PowerMolten Chloride Salt Fast Reactor MCSFR(Elysium Industries, USA)All content included in this section has been provided by, and is reproduced with permission of, Elysium Industries, Unites States of America.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

Elysium Industries, United Statesof America
MSR – Fast Chloride

Coolant/moderator

NaCl-XClv-YClz-UCl3/4PuCl3-FPCly fuel salt / None
(125 / 50), (500 / 200),(1 000 / 400) , (3 000 / 1 200)
Forced circulation

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

Elysium Reactor 500 MWth

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

0.1 + pump head+hydrostatic /slightly higher
650 / 750 (Goal: 950)

Number of fuel assemblies incore
Fuel enrichment (%)

n/a, Full liquid fuel coreCore Discharge Burnup(GWd/tonne)
Purification cycle (months)
Main reactivity control
mechanismApproach to safety systems
Design life (years)

Molten Chloride Salt

10% Pu fissile/(Pu+U total)or ~15% enriched HALEU
300-500 (40-60 years), fertile
added, fuel never damaged
480-720
Fuel expansion in/out of core;
Fertile fuel addition; Passive fueldraining
Passive to Air Cooling

Plant footprint (m2)

Unlimited core, 15-40 forcomponents, 100+ for plant
1/3 size of LWR

RV height/diameter (m)9/4

Seismic design (SSE)

Tension skirt Lateral snubbers

Fuel cycle requirements /ApproachDistinguishing features

U/Pu Closed Fuel Cycle
SNF/DU/NU (1t/GWe-yr)
Fast spectrum, no in-core structure,
60 years fuel life
Conceptual design

Elysium Reactor 3000 MWth
Design status

1. Introduction
The MCSFR is a modular configuration and construction reactor. The MCSFR enables the closing of the fuelcycle, while providing reliable, passively safe, proliferation-resistant, and environmentally-friendly energy(heat/electricity) generation. The fuel is part of the liquid heat transport eutectic fluid with heat directly
deposited in the liquid fuel/coolant.

2. Target Application
The MCSFR is designed for mass production for domestic use and export to address global markets for costcompetitive, low-emission electricity, and high temperature process heat (e.g.: H2, syn-fuel, syn-fertilizer,
desalination, district heating, cement, steel, etc). The MCSFR uses Spent Nuclear Fuel (SNF), plutonium (Pu),or depleted uranium (DU) ‘waste’ with fuel production denaturing and complete consumption (e.g.: US,
Canada, UK, Japan, South Korea, etc.), as fuel sources for internal or export applications and developing
countries with fuel take-back. Utilities can ‘start small’, then add heat exchangers (Hx’s) to increase capacity
without purchasing and licensing a new reactor.

293

3. Design Philosophy
Elysium’s team designed the MCSFR using attributes from:1) water reactors, common, low cost coolant (table salt), fuels (nuclear waste), and qualified materials,2) liquid metal reactors, the low pressure and materials without corrosion concerns,
3) gas reactors, the flow pattern for low containment temperatures, and very high peak temperatures,
4) heat-pipe reactors, passive temperature dependent on/off heat-pipe decay heat removal.
Elysium philosophy includes mitigating public concerns with passive safety, high thermal and fuel efficiency,
SNF/Pu waste denaturing and consumption, dramatically increased fuel supply, elimination of coolant
reactions with water/air/structure, significantly reduced construction, fuel, and operational costs, in addition
reduced proliferation concerns.

4. Main Design Features
Power Conversion System and Cogeneration
The intermediate clean salt heats power conversion unit (PCU) saturated steam to SH steam in the SH, with
35% of SH steam to the steam Rankine turbine-generator (40-50% efficiency) and 65% of SH steam to the
Loeffler boiler.
The use of a salt to SH, i.e. a gas Hx allows use of other process heat gasses, especially at 950oC, on a per Hxbasis to allow flexibility of products. With the high outlet temperatures, process heat applications include H2,synthetic fuel and fertilizer, oil/gas recovery and refining, industrial process heat, and cement manufacturing.
Other applications include district heating/cooling, thermal storage in chloride salt, and desalination. Use ofsalt to SH Hx’s, dramatically reduces water and potential for transient pressures in containment.

Reactor and Core-unit
The core size is minimized to barely maintain criticality with no in-core structures and is near spherical. The
RV and all fuel and intermediate component shells are cooled by cold coolant inside, so includes pipe-in-pipe
nozzles. The core is pure salt, except for a downcomer shroud near the edge of the core, and the lower RV is
the core edge with an ex-RV radial reflector and above core reflector/shield.

500 MWth Elysium Reactor Unit

3000 MWth Elysium Reactor Unit – Axial View

Reactivity Control
Reactivity control is via the negative temperature and void coefficients. As the fuel-salt temperature increases
the fuel-salt expands and fissile/fuel salt is squeezed out of the core, reducing power and vice versa. The longterm reactivity adjustments are made by on-line fertile fuel additions. The reactor can also be shut down bytripping pumps to allow draining to criticality safe/passively cooled drain/expansion tanks.

Reactor Vessel (RV)The RV is low pressure, thin-walled stainless steel, with up to 6 nozzles, cooled by cold fuel salt everywhere
inside, and submerged in a cooling/shield tank of clean salt for corrosion prevention/cooling outside. The RV
near the core is the core outer diameter and is replaceable separately from the upper RV nozzle region. A
reflector, if needed, is outside the RV.

294

3000 MWth Elysium Reactor Unit – Front View

View of the MCSFR Underground Reactor Building

Fuel Characteristics
The fuel-salt is NaCl-XClv-YClz-UCl3/4-PuCl3-FPCly, allowing it to contain ~30% total actinide Chlorides witha 10-20% fissile fraction and 99.9% actinide consumption. Fissile options include: 1) Preferred-Reactor Gradeor Weapons Grade Plutonium (RGPu and WGPu) due to revenue and is denatured with SNF, if needed, 2) highassay low enriched uranium (HALEU), 3) high enriched uranium (HEU) denatured at fuel production site.
Fertile options include: a) Prefered-LWR/CANDU SNF, b) Depleted Uranium (DU), c) Natural Uranium (NU),d) residual U from other mining, coal ash, or seawater extraction, e) thorium (Th) combined with >88% U238
for denaturing. Pu fissile is added for startup only. HALEU requires continued feed-in declining in enrichment
over 5-10 years. Fissile is iso-bred plus enough to counteract fission product poison buildup to prevent needing
purification for many decades.

Nuclear Heat (Steam) Supply System (NHSS)The MCSFR has a fuel/coolant loop, clean intermediate salt loop same as fuel salt, without the fuel/fission
products (FPs), etc., the steam super-heater (SH) loop all inside containment, with a Loeffler steam boiler,
outside of containment. The reactor has 1 to 6 Hx/pump loops for each fuel and clean salt loop. Each of theloops can be used for 6 different heat applications/customers. Loops can operate at variable power up to
maximum with fuel cost not a concern at part-load.

Primary Loop
The primary loop is chloride fuel salt. The arrangement is modular, like an HTGR to allow for low temperature
containment, yet high temperature for efficiency and process heat applications. Six pipe-in-a-pipes connect ina modular arrangement to up to 6 fuel salt to clean intermediate salt Hx’s, with top mounted pumps after the
Hx to prevent motor immersion and heating. Radial dimension is ≤ 4m for road shipping.

Intermediate Loop
The intermediate loop is the same salt as the fuel salt without the actinides or FPs, and is at a higher pressure
than the fuel salt to ensure any leakage is not of fuel salt outward, inward for dilution/shutdown of the core.
The intermediate loop transports heat to the modularly mounted SH Hx, with the intermediate salt pump on
top of one end of the SH. Radial dimension is ≤ 4m for road shipping.

Fuel Cycle and Length ApproachOn-line fertile fuelling is used. Iso-breeding ratio is ~1.014/yr to offset fission product buildup with a doublingtime of 50-60 years with a fuel life of 40-60 years when it is sent to a central facility for partial purification
and recycling of all actinides, carrier salt components, including all chlorine. The fuel consumption rate is ~1
ton U / GWe-yr. Fuel is added every ~8 hrs-7 days depending on the power level by dropping a fertile pebble
into a perforated basket in the drain tank flow path.

Cooling System
Four different cooling loops are included: the fuel salt, intermediate salt, the power/process heat loop, and theheat sink loop. With the high temperature and very low fuel cost, dry cooling is an economic alternative.

Proliferation Considerations
The MCSFR never contains weapons grade materials. Fuel is not removed for 40-60 years. WGPu or HEU isdenatured at the single fuel production facility. Purification at 60 years never separates Pu from U, Cs and Sr.
This low separation of FPs reduces recycling cost. Only relatively short-lived, ~100 years to below U
background, FPs are removed as waste.

295

5. Safety Features
Engineered Safety System Approach and Configuration
Passive features include the large mass and heat capacity of the fuel, intermediate, and tank salt to slow heatup. Pumps constantly fill the primary system from the bottom and can be tripped passively by high or low salt
temperatures, or by the reactor operator. When pumps stop, fuel salt immediately starts draining from the RV
and Hx’s to a passively shut down high neutron and heat leakage drain tank with external tank salt cooling.
The RV and Hx’s are also tank salt cooled.

Decay Heat Removal System
Passive on/off heat pipe decay heat HX’s remove decay heat to air from the tank salt heat mass, whentemperatures exceed the vaporization point, while also shutting down preventing freezing of the fuel salt in thedrain tanks when decay heat decreases to allow faster plant recovery.

Containment Function
The MCSFR has 3 intrinsic and 3 physical containment barriers to leakage: chemical binding, inward leakage,salt freezing, fuel salt RV and HX shells, shield/cooling tank shell, underground low pressure containment
cylinder. Containment is below grade with the top being an airplane shield.

6. Plant Safety and Operational Performances
The MCSFR is designed for passive, fast load/source following. With a closed fuel cycle, no solid fuelmanufacturing costs, and extremely low fertile fuel costs or revenue from SNF, fuel cost is very low. Turbine
bypass is economically viable for fast load following. Consuming more fuel via bypass operation, i.e., more
SNF consumption, is a revenue offset.

7. Instrumentation and Control Systems
The MCSFR operates on load following the by changing the temperature difference. Temperatures decrease as
fertile is consumed and FPs build up, which is a signal to add more fertile to increase fissile breeding.
Temperature, pressure, level, and flow rate are the primary control systems, with significant peak, or minimum
temperature changes passively initiating draining, and average temperature controlling fuel expansion in/outof the core. Elemental, isotopic, and molecular composition are monitored.

8. Plant Layout Arrangement
The plant is a modular configuration, like an HTGR. The RV is in the center, surrounded by short pipe
mounted, i.e., modular, salt to salt, and salt to process heat HX’s radially around the core in a large salt tank.
The reactor is sized to just achieve criticality. Low initial power can use one HX pair, and allow uprating byadding Hx pairs as needed. The reactor is underground for aircraft and security cost reduction.

9. Design and Licensing Status
The conceptual design of pilot fuel production and reactor are in progress.

10. Fuel Cycle Approach
High fissile startup fuel is made in a single fuel production facility and shipped to the new plant. The MCSFR
is a closed fuel cycle requiring only 1 tonne SNF, DU, NU fuel/GW(e)/yr for each plant. Feed-in fuel is added
daily/weekly, depending on power/burnup rate, for 40-60 years. Every 40-60 years the fuel is removed/sent to
one or two recycling facilities. Short-lived FPs are removed. All actinides and long-lived Cs and Sr are left
together, split in two, and sent out to plants as new/replacement fuel. Chlorine is recycled. Breeding ratio (BR)is ~1.014/yr, with a ~50 yr doubling time. Elimination of solid fuel manufacture dramatically reduces fuel andrecycling cost, consumes Pu and SNF waste.

11. Waste Management and Disposal Plan
The waste forms removed are fission gasses (which are allowed to decay), noble metal solid precipitates, 100year FPs and carrier salt anions. These short-lived FPs can be mined for valuable radio- and stable-isotopes,
then are allowed to cool as high temperature salts in surface stored cans cooled by air for ~100 years until
stable.

12. Development Milestones
2018-2021
2020-2027
2027-2030

296

Pre-Conceptual design is near completion. Small scale tests of key concepts areunderway with large scale testing estimated to start in 2020, followed by an Integrated
Systems Test
Design and building a pilot fuel facility and 10-30 MW(t) pilot plant
Licensing/building Commercial unit

Lithium Fluoride Thorium Reactor(Flibe Energy, United States of America)
All content included in this section has been provided by, and is reproduced with permission of, Flibe Energy, Inc., USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

Flibe Energy, Inc., UnitedStates of America
Molten salt reactor

Coolant/moderator

LiF-BeF2-UF4 fuel salt /Graphite
600 / 250

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

Forced circulation

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

Ambient

Fuel enrichment (%)

Not applicable, uses 233U
derived from Thorium
Continuous refuelling from233
U produced in blanket
Negative temperature
coefficient; control rod
insertion
Passive

Refuelling Cycle (months)
Reactivity control mechanism

Approach to safety systems
Distinguishing features

Design status

500 / 650
LiF-BeF2-UF4

Complete consumption ofthorium resource for energy
generation
Conceptual Design

1. Introduction
The lithium-fluoride thorium reactor (LFTR) design by Flibe Energy is a graphite-moderated, thermalspectrum reactor with solutions of liquid fluoride salts containing both fissile and fertile materials. Thermal
power generated from nuclear fission would drive electrical generation in a closed-cycle gas turbine power
conversion system. The objective is to produce electricity at low cost by efficiently consuming thorium.
Mixtures of fluoride salts raised to a sufficient temperature to allow them to liquefy form an ideal medium inwhich nuclear fission reactions can take place. The ionically bonded nature of the salts prevents radiation
damage to the mixture and allows for operation at high temperature yet at essentially ambient pressure.
The high operational temperatures of the fluoride salts (500-700°C) make them excellent candidates forcoupling to a closed-cycle gas turbine power conversion system (PCS). The supercritical carbon dioxide gas
turbine employing the recompression cycle is proposed and can generate electricity at high efficiencies(approximately 45%).
The LFTR design has a two-region core (feed/breed) and utilizes a closed fuel cycle based on thorium. The
reactor vessel incorporates two plena with a central active core region and the outer blanket area, both filledwith fluoride salt. The 232Th in the blanket region is ultimately converted to 233U through neutron capture andbeta decay. The chemical processing system is used to separate and re-introduce the fertile and fissile material
to the two fluorides fuel-salt streams respectively. Utilizing thorium fuel in a thermal neutron spectrum, thereactor can extract almost all the energy content thus assuring practically unlimited thorium resources and theassociated insignificant basic fuel costs.

2. Target Application
Develop a power-generating nuclear reactor that will produce electrical energy at low cost by efficiently
consuming thorium.

297

3. Design Philosophy
The objective of the liquid-fluoride thorium reactor (LFTR) design proposed by Flibe Energy is to develop a
nuclear power plant that will produce electrical energy at low cost. By utilizing thorium fuel in a thermal
neutron spectrum, the reactor can utilize the energy content of the thorium at a very high efficiency, and to apoint where the Earth's thorium resources practically becomes unlimited. The main design principles are (i)inherent safety, with a no meltdown and non-pressurized core; (ii) simplicity, to have an intrinsically stable
and self-regulating design; (iii) fuel efficiency, and (iv) the potentiality to produce far less waste.

4. Main Design Features
Nuclear Steam Supply System
The nuclear heat supply and power conversion system is included in the simplified flow diagram above. It
includes the reactor and primary loop, intermediate loop, power conversion system, and external coolingsystem. The individual systems are described in more details below followed by other system design
descriptions.

Power Conversion System
The function of the PCS is to convert the maximum amount of enthalpy contained in the heated working fluid
into shaft work and to reject the remaining enthalpy to the environment in an acceptable manner. The
supercritical carbon dioxide gas turbine employing the recompression cycle appears to be the best candidate
for coupling to the reactor. The PCS includes four heat exchangers: the gas side of the gas heater, the gas
cooler, and the high-temperature and low-temperature recuperators. It also includes the main turbine, the maincompressor, the recompressor, and the electrical generator. The PCS interfaces with the intermediate loop
through the gas heater, and interfaces with the external cooling system through the gas cooler.

Reactor Core
The reactor vessel functions to hold fuel salt, blanket salt, and moderator material together in such a way tomaintain a critical configuration at the temperatures and thermal power levels required. In addition, it
incorporates reactivity control mechanisms both active and passive. The fuel and blanket salts are kept
separated in two plena integrated into a single structure within the reactor vessel. Fuel salts are directed into
the appropriate channels as it is circulated through the reactor.
The reactor vessel design incorporates several safety functions. In many accident events, a freeze valve, which298forms part of the vessel and primary loop system, melts and allows fuel salt to drain from the primary loop andthe reactor vessel into the drain tank. The separation of the fuel salt from the solid graphite moderator retainedin the reactor vessel, assures that a subcritical configuration can be established in the drain tank.
The internal graphite structures need to be replaceable since they are subject to a fast and thermal neutron fluxthat is greatly in excess of that which will be experienced by the metallic reactor vessel itself, and thereplacement of these graphite structures will enable the reactor vessel to continue to operate and serve its
function.

Reactivity Control
The reactor vessel accommodates passive and active control rod systems which also have important safetyfunctions. The blanket salt held within the reactor vessel is a strong neutron absorber, and a blanket salt leakfrom the reactor vessel could lead to the reduction in the blanket salt inventory contained in the reactor vessel,increasing reactivity by removing a neutron-absorbing medium. To compensate for this introduction of positive
reactivity, a series of control rods that float in the blanket salt and are thus held outside of the core could be
used. An accidental drain of the blanket salt would remove the buoyancy effect of these rods, allowing themto slide down into the core and add negative reactivity to replace and overcome the negative reactivity lostfrom by the drain of the blanket fluid. These rods would be designed to enter the core passively, without any
operator action, in the event of blanket loss. But it is anticipated that there would also be an active drive systempresent that could drive these rods into the core intentionally in order to have a shutdown effect on the reactor.
It would not be possible to start the reactor unless these rods were fully withdrawn from the core due to theirstrong negative reactivity.
An active set of control rods, of a more conventional design, would also be present in the reactor vessel andwould serve a safety function, allowing the operator to control the reactivity level of the reactor. These rods,
which would comprise a smaller and less potent source of negative reactivity, would be clustered near thecentre of the core and provide finer control over reactivity levels. Other possible designs are also considered.

Reactor Pressure Vessel
The reactor vessel shall be constructed from a material that is suitable for accomplishing its functions at the
anticipated temperatures, stresses, and neutron fluxes that will exist during operation. Current evidence points
to a modified form of Hastelloy-N as the suitable construction material.

Primary Loop
The function of the primary loop is to direct fuel salt through the primary heat exchanger (PHX) in normal
operation, where the fuel salt transfers its heat to the coolant salt. The primary pump provides the necessary
forced circulation. The primary loop system includes the primary pump, the PHX (integrated in the reactor
vessel), the bubble injection system, and the fuel salt drain tank and its associated external cooling system.

Intermediate Loop
The intermediate loop transfer heat from the primary loop to the PCS. The intermediate loop system includes
the PHX, the coolant salt pump, the salt side of the gas heater (or intermediate heat exchanger, IHX), thecoolant salt drain tanks, and the pressure relief (blowout) valves. The intermediate loop also isolates theprimary loop from the high pressures of the PCS using pressure relief valves. The isolation is an importantsafety function. In case of a failure in the high-pressure PCS it will prevent the transmittal of high pressureback through the coolant salt to the primary loop. The primary loop is not designed for high pressures andwithout isolation a break in the PCS could cause component rupture and potentially disperse radioactivity into
the containment. In the event of a failure in the gas heater and the pressurization of the intermediate loop, thepressure relief valves allow coolant salt to leave the loop. This deprives the primary loop of cooling capability
and will lead the melting of the freeze valve in the primary loop and the drain of the primary loop fluid contents
into the fuel salt drain tank (also see passive shutdown and heat removal later).
The use of a coolant salt in this loop leads to a compact PHX, also reduces the fuel salt inventory, and thus theamount of fissile material needed for a given power rating.

External Cooling System
The function of the external cooling system is to reject the heat that was not converted to shaft power in the
PCS to the environment in an acceptable manner. The design shall also prevent the transmission of tritium tothe outside environment.

Fuel Characteristics
Thorium fuel is introduced as a tetrafluoride into the blanket salt mixture of the reactor. The blanket salt
surrounds the active ‘core’ region of the reactor and intentionally absorbs neutrons in the thorium, which leads
to the transmutation of the 232Th via nuclear beta decay, first to protactinium-233 (233Pa) and later to 233U. Boththe protactinium and the uranium are chemically removed from the blanket salt mixture and introduced into
the fuel salt mixture in the reactor to fission. The fission products are later chemically removed from the fuelsalt and in some cases separated and purified before final disposition.

299

Fission Product Retention
The integrity of the reactor vessel plays an important role in minimizing radiation hazards by confining
radioactive fluids to the flow channels and volumes defined by the vessel and its internal structures.
Most fission products, including all of those of greatest radiological concern, form stable fluoride salts that are
retained in the overall mixture under all conditions. Fission products gases, whose removal is important froma performance and safety basis, are easily separated from the fluid mixture and allowed to decay to stability ina separate system.

5. Safety Features
Passive Shutdown and Heat Removal
An important safety function is embedded in the primary loop and is activated when the reactor overheats orloses its coolant flow. A freeze valve is integrated into the primary loop that is maintained frozen by an active
coolant system. When this coolant is lost or if the temperature of the system exceeds its cooling capability, thefreeze valve fails open and the fuel salt drains out of the primary loop and out of the reactor vessel into the fuelsalt drain tank. The fuel salt drain tank is integrated with a separate cooling system that is passively connectedto the outside environment and provides the necessary cooling for the fuel salt within it.

Fluoride Salt Characteristics
The fluoride salt mixtures in question have high volumetric heat capacity, comparable to water, and do notundergo vigorous chemical reactions with air or water in contrast to many liquid metals. The components offluoride salt mixtures have both desirable and undesirable aspects, and the two most important are lithium-7
fluoride and beryllium fluoride. The two natural isotopes of lithium must be separated from one another sinceLi6 (7.5% of natural lithium) is far too absorptive of neutrons to be a suitable component of a reactor fluid.
Beryllium fluoride is chemically toxic but has very attractive nuclear and physical properties. The chemical
processing and purification of fluoride salt mixtures typically involves using powerful reactants such as
gaseous fluorine and hydrogen fluoride which are very toxic and reactive. But the fact that fluoride salt
mixtures are processed in a salt form rather than being dissolved into an aqueous solution mitigates issues ofaccidental criticality considerably, since water is an excellent moderator whereas salts are poor.
Fluoride salts, due to their exceptional chemical stability, have the potential to corrode most structural metalalloys, but some alloys have been developed that hold up very well against any corrosive attack. Invariably
these alloys are based on nickel with a variety of other metallic constituents. Fluoride salts moderate neutrons
sufficiently on their own to prevent the formation of a truly fast neutron spectrum, but are still insufficiently
effective to generate a thermal neutron spectrum. Thus, separate moderator materials are necessary for thereactor and graphite has been proven to be very attractive.
Graphite is not wet by the fluoride salts and does not require cladding. If the surface of the graphite is treated
so that small pores are closed, most fission product gases can be excluded from the graphite and overall
performance will be high. Graphite does experience issues from dimensional distortion over time, but thiseffect can be quantified and compensated for in reactor design.

6. Plant Layout Arrangement
The reactor cavity or silo is below grade and contains the primary circuit.

7. Design and Licensing Status
The design is in an early stage of development and licensing activities have not yet been undertaken.

8. Fuel Cycle Approach
The LFTR two-region core facilitates feed and breed and it utilizes a closed fuel cycle based on thorium.

9. Waste Management and Disposal Plan
LFTRs have the potential to produce far less waste than LWRs along the entire fuel cycle and process chain,from ore extraction to nuclear waste storage. LFTR technology can also be used to consume the remaining
fissile material available in spent nuclear fuel stockpiles around the world and to extract and resell many of theother valuable fission byproducts that are currently deemed hazardous waste.

10. Development Milestones
October 2015
July 2018

300

EPRI-funded study of LFTR design published
DOE award announcement for advanced fluorination development work

ThorCon (ThorCon International, UnitedStates of America and Indonesia)
All content included in this section has been provided by and is reproduced with permission of ThorCon International.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginThorCon International, first
deployment in Indonesia

Reactor type

Thermal molten salt reactor

Coolant/moderator

Molten salts / graphite

Thermal/electrical capacity,
MWt/MWe

557 / 250 (per module)Primary circulation

Forced (4 pumps), 1 per loop

NSSS operating pressures(4 loops: fuel salt, clean salt, solar
salt, steam), MPa absolute

1.06 / 1.30 / 0.56 / 25.7

Core Inlet/Outlet fuel-salt
temperature (oC)

560 / 704

Fuel type/assembly array

UF4 / molten salt

Power conversion process

Supercritical steam turbine

Fuel enrichment (%)

Startup 2.3 / makeup 4.95

Discharge burnup (GWd/ton)

145.8

Refuelling cycle (months)
Reactivity control mechanism

12 months to adding fuel48 months to Can changeover
Negative temperature coef.

Approach to safety systems

Intrinsic, passive

Design life (years)

80

Plant footprint (m2)

67x162

Can height/diameter (m)10.3 / 7.8

Can weight (metric ton)

343

Seismic design (SSE)

1.0

Fuel cycle requirements /approachFissile LEU05, or LEU19
with thorium conversion

Distinguishing features

Full passive safety, short
construction time, low cost

Design status

Preliminary design

1. Introduction
ThorCon is a molten salt fission reactor. Unlike all current operating reactors, the fuel is in liquid form. The
ThorCon reactor operates at near atmospheric pressures and is constructed using automated, ship-style steel
plate construction methods.
The top picture shows two pairs of 557 MWe reactor Cans in (blue) silos within the hull wall. Each 250 MWe
power module contains two replaceable reactors in sealed Cans, depicted in red, which are inside the silos.
One Can of each module produces power while the other is in cooldown mode. After four years the cooled
Can is replaced with a fresh Can, the fuel salt transferred to it, and the used Can starts its 4-year cool down.
The second picture above shows two 500 MWe ThorCon power plants and a CanShip exchanging the Can
containing the reactor vessel and radioactive primary loop. The yellow rectangles are hatches for access bygantry cranes. The middle graphic shows the air cooling towers, fission island, heat exchangers, steam turbinegenerator, and switchgear. The bottom graphic also shows the steam generation cell and secondary heatexchanger cells, flooded in water used as a backup decay heat sink.

301

2. Target Application
The first planned application of ThorCon reactors is to generate electric power in developing nations withfragile grids, so ThorCon is capable of demand discontinuities and black start without grid power. Capital costand generated electricity costs are critical in these markets. ThorCon is cheaper than coal and deploys rapidly.

3. Design Philosophy
Walkaway Safe – If the reactor overheats for any reason, it will automatically shut itself down, drain the fuelfrom the primary loop, and passively remove the decay heat. Operators cannot prevent draining and cooling.
ThorCon has three gas tight barriers between the fuel salt and the environment. In a primary loop rupture, there
is no coolant phase change and no dispersal energy..
Ready to Go – The ThorCon design needs no new technology. ThorCon is a scale-up of the successful Molten
Salt Reactor Experiment (MSRE). A full-scale dual 250 MWe ThorCon prototype can be operating under testin 2026 and then subjected to failure testing before commercial production can begin.
Rapidly Deployable – The entire ThorCon plant is designed to be manufactured in blocks in a shipyard. The
150 to 500 ton, fully outfitted, pre-tested blocks are assembled into a hull containing the complete power plant,
towed to a customer site, and firmly settled in 5-10 m of water. A single large reactor yard can turn out twenty
500 MWe ThorCons per year. ThorCon is more than a power plant; it is a system for building power plants.
Fixable – No complex repairs will be attempted on site. Hatches and cranes permit components of the fission
island to be replaced. The primary loop is totally contained within a Can. Every four years the Can is changed
out, returned to a centralized recycling facility, decontaminated, disassembled, inspected, and refurbished.
Cheaper than Coal – ThorCon requires far fewer resources than a coal plant. Assuming efficient, evidencebased regulation, ThorCon will produce clean, reliable, carbon-free electricity at less than the cost of coal.

4. Main Design Features(a) Power Conversion
Each power module employs four loops
for converting fission heat to electric
power: the primary fuel salt loop inside
the Can, the secondary salt loop, a solar
salt loop, and a supercritical steam loop.
The solar salt loop captures any tritiumthat has made it past the secondary loop,
and more importantly ensures that arupture in the steam generator creates nonasty chemicals and harmlessly vents to
the Steam Generating Cell via an open
standpipe.

ThorCon is a high temperature reactor, with a thermal efficiency of 46.5% using 30oC
seawater compared to about 32% for a standard light water reactor. This reduces
capital costs and cuts cooling water requirements by 45%. It also allows ThorCon to
use the same steam cycle as a modern coal plant.
The fuel salt is a mixture of sodium, beryllium, and uranium fluorides at 704oC. The(red) Can contains the (orange) reactor called the Pot. The primary loop pump pushes
the fuel salt at 3300 kg/s through the (blue) primary loop heat exchanger (PHX).
The PHX transfers heat to secondary salt in (green) piping above the Can. The 560oC
fuel salt is piped back to the bottom of the Pot, where the graphite moderator slows
neutrons, which fission uranium in the fuel salt as it rises through the Pot, heating thesalt. Neutrons also convert some fertile 238U to fissile fuel.
The steel silo cold wall shown in dark blue is cooled by surrounding water. The Can
is cooled by thermal radiation to the silo cold wall. The heated water rises by natural
circulation to above-deck, air-cooled radiators, where it is cooled and returns to thebottom of the cooling wall via the basement.

302

A loop of hot secondary salt of mixed NaF and BeF2 is pumped through the PHX in the Can to a Secondary
Heat Exchanger (SHX). It transfers heat to a third loop of NaNO3/KNO3, called solar salt because of its use
in thermal solar power plants. The solar salt, shown in purple, transfers heat to a supercritical steam generator
and steam reheater.
All radioactive fission products and fissile material
are contained within the Consolidated Boundary
indicated by the red outline. The space is shielded bysteel sandwich walls protecting the crew andprevents intrusion. This space is guarded by IAEA
seals and monitoring.

(b) Reactor Core
The reactor core is inside the pot. The core is 90%
filled with graphite logs that moderate neutronenergies. The core is 6.5 m in diameter and 4 m high.

(c) Reactivity Control
Reactivity is controlled by fuel salt temperature,
which rises as steam turbine power generation heat
demand drops. Fission decreases as salt, graphite,
and Pot temperatures rise. Temperature adjustment
rods compensate for neutron absorption by xenon
during power changes.

(d) Reactor Pressure Vessel and Internals
The Pot reactor vessel pressure is 0.33 MPa, insufficient to spread radioactive material into the environment
in an accident, so it does not have the central safety importance that it does in an LWR. The Pot contains
circulating fuel salt, graphite moderator, shutdown rod channels, and temperature adjustment rods.

(e) Fuel Characteristics
The fuel salt is NaF-BeF2-UF4 72/16/12. As fissile U is consumed fissile Pu239 is generated, but not enough to
replace the fuel burned. Makeup fuel must be added continuously.

(f) Fission Product Management
Fission product off-gases include xenon and krypton, mixing into helium cover gas flowing over the fuel salt
surface in the Pot header tank. The off-gas mix flows through tanks inside the Can to allow most of theradioactive gases to decay to solid fission products which are captured within. The cooled off-gas flows to
holdup tanks and charcoal beds in the SHX to allow the longer lived radioactive noble gas to decay except Kr85 and tritium. The xenon and krypton are removed to tanks and the helium recycled.

5. Safety Features(a) Engineered Safety System Approach and Configuration
The ThorCon negative temperature coefficient provides passive temperature stability. The large margin
between the operating temperature of 704oC and the fuel salt boiling temperature of 1430oC exceeds possible
temperature excursions, so radioactive material cannot be vaporized.

(b) Decay Heat Removal System
The FDTs radiate fuel salt decay heat to steel jackets cooled by the same naturally circulating water circuit
used to cool the cold wall, passively removing the decay heat to the air cooling towers. No operator intervention
is needed at any time. No electric power is necessary, though fans keep basement water cooler. Cooling using
natural air flow continues indefinitely.

(c) Emergency Core Cooling System
If the air-cooled radiators in both cooling towers become disabled, basement water provides passive coolingfor at least 40 days. Cooling time can be extended indefinitely by supplying more water from sources such as
the desalination system, domestic water, ballast tank water, or external water via on-deck fittings.

(d) Containment System
The primary loop is totally contained within the removeable Can and FDT. This system constitutes primarycontainment for fuel salt during power operations and accidents from power operations. The primarycontainment for spent fuel during transfer and storage are double walled piping and tanks, also within the
consolidated boundary secondary containment. These first two containments are gas tight. Tertiary
containment is provided by the ship hull structure of three meters of concrete sandwiched between 25 mm
steel plates.
303

(e) Chemical Control
Makeup UF4 fuel enriched to 5% 235U is added continuously to increase reactivity. Fertile 238UF4 can be added
to decrease reactivity. Beryllium metal is added to maintain proper redox potential as fluorine is freed via UF4
fission. No boron additions occur.

(f) Spent Fuel Cooling Safety Approach/System
ThorCon has two gas tight barriers and one containment barrier between the fuel salt and the environment.
ThorCon reactor operates at near-ambient pressure. In the event of a primary loop rupture, there is little
dispersal energy and no phase change and no vigorous chemical reactions (like zirconium and steam).

(g) Spent Fuel Salt
Operating with uranium fuel in feed-bleed mode, continuous addition of makeup fuel salt causes removal of
used fuel salt from the primary loop to an overflow tank in the Can. Periodically, fuel salt in the overflow tankis transferred to a vault tank. The removed, cooling fuel salt is as well protected as the fuel salt being fissioned.
Spent fuel salt may later be transferred to a shipping cask and transferred to the visiting CanShip and shipped
to a fuel salt handling facility for future recycling.

6. Plant Safety and Operational Performances
Load following relies on reactor physics. As electric power demand decreases, less heat is removed from thefuel salt by the PHX and reactivity and power decrease as temperature rises.

7. Instrumentation and Control Systems
Instrumentation and control systems are not safety critical for ThorCon. Argonne National Lab is adapting its
isotopic concentration sensors to monitor ThorCon fuel salt components. Numerous sensors will record andreport the condition of power generation..

8. Plant Layout Arrangement
Two 557 MWt power modules drive a single 500 MWe turbine/generator. This allows using competitively
priced, efficient supercritical steam turbine-generators.

9. Testing Conducted for Design Verification and Licensing
Measurements of molten salt properties are being conducted at independent laboratories. Soil testing has been
completed in the bay of the island site for the Indonesia demonstration plant. The non-fission test platform isbeing designed to test components such as molten salts, sensors, pumps, valves, pipes, vessels, graphite
moderators, and heat exchangers at operating temperatures and pressures, using externally supplied electric
power rather than fission power. This facility will be used to verify safety systems including performance forcontainment and cooling during severe accidents. Neutronics testing will wait for the demonstration plant.

10. Design and Licensing Status
Preliminary design is complete. Some detailed designs are being discussed with specialty component suppliers.
License discussions are continuing with the Indonesian regulator, Bapeten, along with university professors
and government officials.

11. Fuel Cycle Approach
Fuelled by 5% enriched uranium additions, ThorCon can operate in feed and bleed mode, continually feedingin fresh fuel salt and bleeding excess volume to the onboard storage vault tanks. Once HALEU becomes
available in power plant level quantities, a 500 MWe ThorCon can operate as a thorium converter using 5.3
kg of 19.7% enriched uranium and 9.0 kg of thorium per day, on average. After each 8-year fuel cycle, theused fuel salt is transferred to the vault tanks, which have capacity for 80 years of this operation.

12. Waste Management and Disposal Plan
Spent fuel salt is transferred to passively cooled vault tanks in the SHX cell for storage up to 80 years. Onceappropriately cooled, the fuel salt can be transferred to a shipping cask and removed by crane and loaded to a
CanShip to transfer it for reconditioning or final storage.

13. Development Milestones
2022
2023/242025/26
2027
2028

304

Pre-licensing vendor design review in Indonesia; Preliminary design complete
Construction of Non-fission Test Platform; Testing of the Non-fission Test Platform
Construction of the demonstration power plant; Begin testing of the demonstration power
plant
Complete testing of the demonstration power plant and obtain type license
Begin commercial construction of multiple power plants; Start of commercial operation

PART V. MICROREACTORS

305

306

Energy Well (Centrum výzkumu Řež s.r.o.,Czech Republic)
All content included in this section has been provided by, and is reproduced with permission of, Centrum výzkumu Řež, Czech Republic .

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
Centrum výzkumu Řežof origin
Czech Republic
Reactor type
Fluoride high temperature reactor
Coolant/moderator
Molten salt FLiBe
Thermal/electrical capacity,
20 / 8MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
AtmosphericCore inlet/outlet coolant
temperature (oC)

650 / 700

Fuel type/assembly array
Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)

TRISO
25

Reactivity control mechanism
Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

Control rods
Active/passive
Not defined< 40007 / 3.5< 100
Yes
Once throughDistinguishing features
Design status

High passive safety features
Pre-conceptual design

15
84
70

1. Introduction
Energy Well is a Fluoride High temperature micro Reactor of 20 MW(t) under development using uniqueknowledge in the Czech Republic on molten salt technologies. The project is under grants from the Ministryof Industry and Trade. The purpose of the project is to develop an advanced, inherently safe low-power hightemperature reactor. The design is mainly intended for remote areas as a long-term source of electrical energy
and heat for island networks. Therefore, the reactor and associated power plant and/or a heat plant must meetthe following requirements, among others: 25 MW(t) maximum power; transportable; long fuel cycle; fuelenrichment < 20%; and thermal efficiency > 40%.

2. Target Application
Energy Well is focusing on operation both in remote and in populated areas, focusing on production ofelectricity, heat and hydrogen as a means of energy storage. The purpose is to provide a clean stable energy
source in synergy with the large-scale nuclear power reactors, heating plants and the renewable sources ofenergy.

3. Design Philosophy
The 20 MW(t) Energy Well adopts a 7-year fuel cycle, low power density, high use of passive safety andsimplicity. The design allows a modular approach and requires minimum on-site welding operations.

307

4. Main Design Features
Power Conversion
The power plant includes threecooling circuits. Liquid fluoride
salts are used as a heat transfer
medium (FLiBe, NaBF4) in the
primary and secondary circuits.
Carbon dioxide in a supercriticalstate (sCO2) is used in the tertiary
circuit. The tertiary circuit considers
an Ericsson–Brayton-based cycle
optimized
configurationfortransformation of the heat to electric
power. The primary circuit removes
the heat generated in the core of thereactor, while the secondary circuit
separates the active primary and thehigh-pressure tertiary circuit while
ensuring the heat transfer.

Process Flow Diagram of Energy Well reactor with simple heat recovery system

Reactor Core
The reactor core includes, 25 hexagonalfuel assemblies, 6 central graphite blocs, anexternal graphite reflector and 19 control
rods. The core is 2 m high and has adiameter of 3 m with the reflector. The
current core arrangement is shown in the
figure on the right.

Fuel Characteristics
The Energy Well core fuel assemblies are
made of TRISO spherical fuel particles thatare blended together to form fuel compacts.
TRISO coated fuel particles are composedof a uranium kernel coated in successive
layers of pyrolytic carbon, silicon carbide,and an outer layer of pyrolytic carbon. The
fuel compacts are stacked in a drilled
hexagonal graphite block that also contains
holes for molten salt coolant flow. A fuelenrichment of 15% has been selected.

Core Arrangement

Reactivity Control
The primary safety system ensuring reactivity control are the control rods with a set scram signal. Scram is
activated when the neutron flux or other technological parameters in the core are increased. During normal
operation, control rods are kept in operational position by magnets that are power fed. In case of deviationfrom normal operation parameters, the Limiting System (LS) regulates the reactor power. A secondary
independent shut-down system is currently being studied and shall be implemented in the design.

Reactor Pressure Vessel and Internals
Energy Well is a pool type reactor with molten salt FLiBe as the primary coolant. The primary cooling circuit
includes the following main parts: a reactor core with top mounted control rods, a reactor vessel with coverlid, a graphite reflector, a core supporting plate, a flow skirting, six primary heat exchangers molten salt/molten
salt and main circulation pumps.
The molten FLiBe flows upward through the core and then, guided by the flow skirting enters the main
circulation pumps that circulate the molten salt to the six primary heat exchangers located on the periphery of
the reactor vessel above the core. At the outlet of the heat exchangers the molten salt flows in a annular space
between the flow skirt and the reactor vessel downward through the core supporting plate.After passing
through the core supporting plate, the FLiBe is directed upward to enter the bottom of the core.

Reactor Coolant System
The primary coolant is a molten fluoride salt containing lithium and beryllium (Li2BeF4— referred to as“FLiBe”). It has a melting point of 459°C, a boiling point of 1430°C, and a density of 1.94 g/cm3 . The heat
capacity of Flibe is 4540 kJ/m3 , which is similar to that of water, more than four times that of sodium, and308

more than 200 times that of helium (at typical reactor conditions). There is also considerable experience with
FLiBe in nuclear systems since it was used in both the primary and secondary loops of the Molten Salt Reactor
Experiment (MSRE) and related test loops. The relatively high melting point of FLiBe will require special
design features. Experience feedback from sodium-cooled fast reactor, lead cooled submarine reactor and
MSRE is being considered during the Energy Well development. The use a pool-type reactor vessel shall
reduce some of the challenges related to molten salt freezing in the primary circuit. Molten salts are alsotransparent which is an advantage during refuelling, maintenance operations and inspection.

Secondary System
The secondary circuit physically separates the primary circuit from the tertiary circuit and creates a pressure
barrier in case of leaks in the salt/salt exchanger between the primary and the secondary circuit. The secondary
circuit includes 3 main components: a salt/salt heat exchanger; a salt/sCO2 heat exchanger; and a circulation
pump. The heat exchangers create the interface of circuits, and the pump ensures the required mass flow of
secondary salt to recover the thermal power out of the primary circuit. The secondary circuit is equipped with
auxiliary systems that include molten salt refilling, salt purification system, expansion volume to cope withmolten salt volume change. The solidification temperature of the salt is a of key parameter. Therefore, the inlet
temperature of sCO2 shall include a margin to avoid solidification of the salt. The ‘NaBF4’ salt is used for thesecondary circuit with 384°C solidification temperature. Alternatively, LiF-BeF2 with a solidification
temperature of 455°C and FLiNaK with a solidification temperature of 454°C, could be foreseen.

Steam Generator
The function of tertiary circuit is for the conversion of heat to electric energy with the sCO2 as working fluid.
The sCO2 technology was identified as the most compatible with this type of SMR. The main benefit is ahigher thermodynamic efficiency due to high temperature through compression close to the critical point (7.38
MPa, 30.98°C), or even in the liquid phase. As a result, the requirements of the compressor are lower, and theintercooling is not required. On the other hand, the limitations of the sCO2 circuits lie in their technical‘immaturity’. The change of media properties significantly complicates the design of components. A range ofthe experimental facilities with a power between 100 kW and 10 MW are being built around the world in orderto verify this technology.
There are dozens of possible sCO2 cycles layouts. Based on preliminary studies, the recompression Brayton
cycle with heat regeneration was selected for the Energy Well system as reasonable compromise between thecomplexity and efficiency of the cycle. Relatively high thermodynamic efficiency of the cycle of 42.71% wasreached for the nominal operational conditions.

5. Safety Features
Engineered Safety System Approach and Configuration
The design of Energy Well reactor has a high focus on passive safety and simplicity. The main safety featuresof the reactor include: Atmospheric pressure in primary and secondary circuits; primary circuit is underground;
low power density; and use of natural circulation and passive safety systems.

Decay Heat Removal System / Reactor Cooling Philosophy
Passive residual heat removal from the primary circuit through the reactor vessel is adopted in case of loss offlow. In the initial phase of event, the high heat capacity of molten salt will allow the heat output to accumulate
in the primary coolant until the heat losses exceed the residual power. Maximum fuel temperature reaches
730°C.

Spent Fuel Cooling Safety Approach / System
The spent fuel cooling system is still to be defined.

Containment System
The TRISO fuel envelope is the first barrier to prevent the spread of fission products. Depending on the design,
the fuel assembly could be considered as a barrier since the TRISO fuel is enclosed in a thick graphite matrix.
The reactor vessel, the containment vessel, the pit (together with the maintenance room shielding ceiling) andthe reactor building are additional barriers of the containment system.

Chemical Control
The chemical control of the molten salts in the primary and secondary circuit is still to be defined. Feedback
experience from MSRE operation shall be considered.

6. Plant Safety and Operational Performances
Air or water are foreseen as heat sink for the facility. The facility is foreseen to be operated at constant powerwith a refuelling performed on-site every 7 years The refuelling is performed by a fuel handling machine
similar to the one used at the Fort Saint Vrain reactor. Fuel handling equipment is designed to be portable andwill not be left on site when not needed to prevent access to fissile material.
309

The main circulation pumps and the primary heat exchangers connecting flanges are bolted to the reactor vessellid. The limited size of the primary circuit equipment and the location of the primary circuit connecting flanges
above the molten salt level shall facilitate the maintenance operations.
To target markets with a demanding power up to 100/200 MW(t) a multiple unit approach is foreseen.

7. Instrumentation and Control System
The I&C system is currently under development and being tested on the experimental loops currently operatingin the Centrum výzkumu Řež s.r.o.

8. Plant Layout Arrangement
The Energy Well reactor building layout
is shown in the figure on the right. The
primary, secondary and tertiary circuit
are located in a common steel building.
The three circuits are located in separate
rooms with separate transfer doors. The
reactor vessel with the primary circuit islocated in a concrete pit in the reactor
room. Additional pits can be added in
the reactor room to implement morereactor units and increase the delivered
power. The reactor room is equippedwith an overhead crane to handle
shielding lids, transfer casks andequipment. Control rods actuators andmain circulation pumps motors arelocated in a maintenance room above thereactor vessel and below the groundlevel. This room can be inerted during
maintenance operations and refuellingto prevent air ingress in the reactor vessel.

3D view of Energy Well reactor building

9. Testing Conducted for Design Verification and Validation
Material corrosion test facilities are built and currently in operation in Centrum výzkumu Řež. A forced
circulation loop with molten FLiBe is currently being built to verify thermohydraulic parameters.

10. Design and Licensing Status
As of 2022 the Centrum výzkumu Řež is preparing a basic design of the Energy Well reactor including
experimental tests. Build an integral testing facility is planned. The Centrum výzkumu Řež is in close
coordination with the State office for Nuclear Safety.

11. Fuel Cycle Approach
A once-through scheme approach is foreseen for the fuel. Once the 84 months fuel cycle accomplished, thespent fuel is removed from the core and transfer to the on-site spent fuel storage area. Fresh fuel is shipped tothe site, inspected and then installed in the reactor vessel with the fuel handling machine.

12. Waste Management and Disposal Plan
The waste management and disposal plan are under development.

13. Development Milestones
2010 20172017 - 2022
2022 - 2027
2027 - 2032
2032 - 2040

310

Basic research in the neutronics, thermohydraulics and material compatibility in regard
to the FLiBe salt
Pre-conceptual design phase and technology validation
Basic design
Experimental verification using an integral test facility
Finalization for manufacturing

MoveluX (Toshiba Energy Systems &Solutions Corporation, Japan)
All content included in this section has been provided by, and is reproduced with permission of, Toshiba Energy Systems & Solutions Corporation,
Japan.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
Toshiba Energy Systems & Solutions
country of origin
Corporation, Japan
Reactor type
Coolant/moderator
Thermal/electrical
capacity, MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary),
MPaCore inlet/outlet coolant
temperature (oC)
Fuel type/assembly array
Number of fuelassemblies in the core
Fuel enrichment (%)
Refuelling cycle(months)
Core discharge burnup(GWd/ton)
Reactivity control
mechanismApproach to safetysystems
Design life (years)
Plant footprint (m2)
RPV height/diameter(m)
RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approachDistinguishing features

Design status

Heat-Pipe cooled and calcium-hydride
moderated reactor
None (Sodium heat-pipe cooled) / Calcium
hydride (CaH2)
10 / 3 – 4
Natural
0.1 /0.3
Heat pipe: 680 / 685
Heat Exchanger: 450 / 680
Silicide (U3Si2) / Hexagonal
66 (fuel), 72 (CaH2)
4.8 – 5.0
Continuous
1.0
In-Ga Expansion Module (IGEM)
Active / Passive
10 – 15100
2.0 / 6.0TBE
0.3 g
Either once-through or closed fuel cycle
scheme depending on country’s fuel cycle
policyUse inherent passive natural principles forreactor shut down by moderator material
property and decay heat removal from thesurface of the reactor vessel by natural
circulation of air
Conceptual design

1. Introduction
MoveluX, Mobile-Very-small reactor for Local Utility in X-mark, is a 10 MW(t) class multi-purpose micro
reactor. A heat-pipe is used as a primary core cooling that provides passive safety as well as system
simplification. MoveluX uses low enriched uranium fuel of less than 4.99 wt% that improves nuclear security
and non-proliferation. Moderator material is required to reduce the core size. In addition, high temperature
operation is essential for the multi-purpose micro reactor. Therefore, calcium-hydride (CaH2) capable ofoperating at up to 800°C is adopted for the moderator material.

2. Target Application
The MoveluX reactor system is a multipurpose energy source that can be used to produce electricity, hydrogenand high temperature heat. Since the reactor system can be installed in remote sites, the heat can be provided
for chemical plants and steel mills. When it is used as a power plant it can be used as a base load power sourceon small grids, possible in combination with renewable energy sources. Since MoveluX generates around 3 –311

4 MW(e) and it can also be used for off-grid applications in remote places.

3. Design Philosophy
The MoveluX reactor system is designed as a multi-purpose energy source that includes off-grid and microgrid electricity, high temperature heat source, hydrogen production and so forth. Designed to produce 10MW(t), it can be used also for electricity production, possibly varying the output. Therefore, when theMoveluX reactor system is connected to a small/micro grid with load following as required and achieved by
the passive reactivity control system. The figure below shows the MoveluX reactor system as a multi-purpose
energy source in such a micro grid.
The major provisions of the MoveluX reactor for resource use optimization are as follows:- Simplified plant design contributes to wastes reduction during operation and decommissioning.
- Low maintenance requirement using no-moving parts component contributes to low maintenancecosts/labours and low waste amount.
- Reduced emergency planning zone contributes to accident management burden and/or cost such as forevacuation.

Utilization image of the MoveluX

4. Main Design Features
Power Conversion
The MoveluX is considering using the gas turbine system for power conversion. The MoveluX can provide ahigh temperature of around 680°C, therefore, the gas turbine was selected from the viewpoint of the conversion
efficiency.

Reactor Core
The MoveluX core consists of the fuel, moderator, heat-pipe and control devices. In this core, uranium silicideand calcium-hydride were used as a fuel and moderator materials. The maximum fuel enrichment is set as 4.99
wt%, from the viewpoint of economics and non-proliferation. Fuel components are installed to the core in the
fabrication phase and loaded as a lifetime core, i.e. this fuel will not be extracted from the core and therefore
fuel handling is not required during operation. The fuel and moderator has 10 cm wide-hexagonal shape except
for the sides where provision is made (cut off is made) for the heat-pipe installation.

Fuel Characteristics
The Uranium silicide fuel has high melting point and large thermal conductivity.

Reactivity Control
A safety-rod is placed at the centre of the core for criticality safety assurance and reactor start-up. The core
reactivity is controlled by the passive reactivity control device, such as In-Gd Expansion Module (IGEM).
In the case of emergency such as loss of cooling ability and/or RIA (with a postulated failure of the safety rod
insertion) the core will be shut down autonomously by the material property of calcium-hydride. Specifically,
calcium-hydride ability to moderate neutrons (neutron moderation power) decreases with increasing
temperature since hydrogen dissociates in high temperature environments above 800°C. The proposed core
design therefore has inherent safety for criticality in emergencies. If the hydrogen pressure is increased by its
dissociation, hydrogen will be processed at a hydrogen processor (details still to be examined).
312

Reactor Pressure Vessel and Internals
Reactor vessel with atmospheric pressure is adopted.

Reactor Coolant System
The primary circuit of the MoveluX reactor system makes use of heat-pipes which is one of the passive coolingdevices, therefore, the primary circuit does not have pumps or other forced circulation devices. In the current
design, sodium is selected as a working fluid of the heat-pipe from the viewpoint of usable temperature andheat transportability. The pressure in the primary system can be set close to atmospheric pressure since theproposed system does not utilize a pump for primary fluid circulation. Therefore, the risk of large-scale
radioisotopes release can be reduced.

Secondary System
The secondary side of the MoveluX reactor system is currently a helium gas system. This gas system canprovide high temperature around 700°C and therefore usable not only for electric power generation, but also
for heat supply, hydrogen production and so on. For electricity generation a Brayton cycle can be used as thepower generation system.

Steam Generator
The MoveluX not uses the steam generator in the current design, however, this option can be examined.

5. Safety Features
Engineered Safety System Approach and Configuration
The reactor vessel is place below ground to enhance reactor protection and radiation shielding. The radiation
shield on top of the reactor is protect the reactor from physical attacks. For the natural hazards, the MoveluX
reactor protection is by the core inherent safety characteristics based on the natural principles.

Decay Heat Removal System / Reactor Cooling Philosophy
The decay heat after reactor shut down is removed passively. If heat-pipe and secondary circuit are keeping
their function, decay heat removal are the same as for heat removal during nominal operation. In the case ofloss of cooling ability of the heat-pipe, the decay heat is removed from the surface of reactor vessel by natural
circulation of air. Furthermore, in the reactor vessel, the heat is transported from the centre to the periphery of
the core by natural circulation and thermal conduction of liquid Pb-Sn, present in the gap between the fuel,heat-pipe and moderator. This decay heat removal system does not require a power source and can therefore
realize a long (infinite) grace period. Consequently, the risk of core meltdown is expected to be very small.

Containment System
The fuel material is contained in the reactor vessel. The fuel material is separated from primary cooling system,
because, heat-pipe is closed heat transportation device. Additionally, the heat-exchanger between heat-pipeand secondary circuit is functioning as one of the boundaries. Therefore, radionuclide will be confined in thereactor vessel unless reactor vessel breaking.

Chemical Control
The MoveluX reactor does not required chemical control.

Schematic view of the MoveluX reactor system313

6. Plant Safety and Operational Performances
For the MoveluX reactor system operation, water as a coolant is not required in current design, because thefinal heat sink is assumed to be the atmosphere. The reactor operation will be automated as possible by passive
control devices based on natural principles. Thus, manpower for the reactor operation is minimized. Because
remote monitoring and operation would become an option for operation cost reducing.
Core Damage Frequency (CDF) : Extremely low.
Refueling : No refueling during the plant life time.

7. Instrumentation and Control System
Few I&C devices are installed to the MoveluX reactor system for the reactor start-up, monitoring and active
control. Technically, the manned operation is not required during nominal operation in current design concept.

8. Testing Conducted for Design Verification and Validation
To be developed.

9. Design and Licensing Status
MoveluX is at the conceptual design stage.

10. Fuel Cycle Approach
The MoveluX reactor can be applied to either once-through fuel cycle scheme or closed fuel cycle scheme. It
mainly depends on user country’s fuel cycle policy.

11. Waste Management and Disposal Plan
The reactor vessel contains the spent fuel as it is, additionally, this spent fuel is carried out to temporary storagesite with reactor vessel. After that, treatment of the spent fuel depends on the country’s fuel cycle policy. On
one hand, in the once-through scheme, spent fuel is extracted from the core at the facility, then, spent fuel isstored to the cask for disposal. On the other hand, in the closed fuel cycle scheme, spent fuel is re-processed
and re-fabricated as a fresh fuel for recycling use in MoveluX, LWR or FR.

12. Development Milestones
2015
2017
2019
2028
2030
2035~

314

Start fundamental study based on the space reactor design
Complete reactor type decision
Start concept design
Complete concept design and component demonstration
Complete system demonstration without nuclear fuel
FOAK demonstration

ELENA (NRC “Kurchatov Institute”,
Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of the NRC "Kurchatov Institute", Russian Federation

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

National Research Centre"Kurchatov Institute" (RRC
KI), Russian Federation
PWR

Coolant/moderator

Light water / light water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

3.3 / 0.068

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

19.6 / 0.36

Number of fuel assemblies in thecore
Fuel enrichment (%)

109

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

57 600 / 27 390

Reactivity control mechanism

Control rods and absorber rods

Approach to safety systems

Passive

RPV height/diameter (m)3.7 / 1.25

Seismic Design (SSE)

VIII (MSK-64)

Fuel cycle requirements /ApproachDistinguishing features

Initial factory load for theentire lifetime
25 years without refuelling,
passive reactivity control andunattended operation
Conceptual design

Design status

Natural circulation

311 / 328
UO2 pellet; MOX is an option

15.2

300

1. Introduction
The ELENA nuclear thermoelectric plant (NTEP) is a direct conversion water-cooled reactor without on-siterefuelling capable of supplying 68 kW(e) of electricity and 3.3 MW(t) of heating capacity for 25 years without
refuelling. The technology and techniques were developed incorporating experience from the construction andoperation of the GAMMA reactor for marine and space application. The ELENA NTEP is designed as an"unattended" nuclear power plant (NPP), requiring nearly no operating or maintenance personnel over the
lifetime of the unit. The conceptual design was developed by the National Research Centre “Kurchatov
Institute” (NRC KI). The ELENA NTEP is a land-based plant; however, in principle it is also possible todevelop underground or underwater versions. The reactor and its main systems are assembled from factoryfabricated finished components or modules, whose weight and dimensions enable any transport delivery
method for the complete plant, including helicopter and ship. The specific features of the design include
capability of power operation without personnel involvement, compensation of burn-up reactivity swing andother external reactivity disturbances without moving the control rods and the use of thermoelectric energy
conversion to produce electricity.

2. Target Application
The unattended ELENA NTEP plant is designed to produce heat for towns with a population of 1500–2000
and located in remote areas where district heating is required. Since it is auxiliary in nature, the electricity
generation of 68 kW could be used for the in-house power needs of the plant and to supply electricity toconsumers requiring a highly reliable power supply, such as hospitals, etc. A desalination unit can be used in
combination with the ELENA NTEP.

315

3. Design Philosophy
The ELENA reactor is designed with an integrated primary circuit. The design features of ELENA ensure highreliability and safety, eliminate adverse environmental impacts, and make the ELENA NPP an attractive sourceof heat and power supply for small settlements located in remote areas, including seismic and draught areas,
as well as in uninhabited or underwater stations, e.g., robotized systems for investigation and extraction ofocean resources or hydrology research laboratories.

4. Main Design Features
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) consists of a reactor core internals and steam generators. The designis based on an integral reactor located in a large volume of secondary water. The NSSS is enclosed in acylindrical vessel that is embedded in a reactor pool structure which is filled with water. Electric power isgenerated in semiconductor thermal battery due to the temperature difference provided between primary andsecondary circuits.

Reactor Core
Pellet type uranium dioxide fuel is used with the average 235U enrichment of 15.2%; the neutron moderator
and coolant is water specially treated according to specified water chemistry. Cylindrical fuel elements withstainless steel cladding are installed in 109 fuel assemblies of 55 fuel elements each; 216 absorber rods withboron carbide based neutron absorber are divided into 6 groups. Fuel assemblies also include burnable
absorbers made of Gd-Nb-Zr alloy. The 235U load is 147 kg.

Reactivity Control
A reliable operation and reactivity control are achieved through the implementation of passive reactivity
regulation and control systems. The control and safety systems, including the control rods and control rod drive
mechanisms are used for reactivity control. The control and safety systems are designed to be fail safe. The
ELENA reactor target is to provide a small total reactivity margin in a hot core so as to secure the survival ofan unprotected transient overpower with no core damage. It also ensures reactivity self-regulation throughout
a very long period of unattended operation.

Reactor Pressure Vessel and Internals
The cylindrical core with a height of 850 mm and an equivalent diameter of 833 mm is installed in a steel shellwith a diameter of 920 mm and encircled by an iron-water shield. The strengthened stainless steel reactor
vessel has an internal diameter of 1250 mm and a height of 3700 mm with a wall thickness of 132 mm.

Reactor Coolant System
The ELENA reactor is a naturally circulated primary system with an integrated reactor coolant system. The
complete reactor system is fabricated from stainless steel. Natural circulation of coolant in both circuits ensures
the NPP is capable of unattended operation without on-site refuelling for up to 25 years. The temperature ofwater within the third loop is about 100oC. The power level is primarily dependent upon the temperature of
the third loop. The internal space for heat transport to consumers is connected to an air-cooled heat exchanger
enclosed in the draft tube for excess heat discharge to the atmosphere.

Pressurizer
The ELENA has three water coolant loops. The primary coolant loop is completely contained within thesecondary barrier. Heat is transported from the core to the consumer though a four-circuit system:- The primary circuit (circuit I) with natural circulation of the coolant (water with a pressure of 19.6 MPa)transports heat from the core to the thermoelectric generator (TEG) modules cooled by the circuit II coolant(water with a pressure of 0.36 MPa);- Circuit II (intermediate circuit) removes heat from the cold joints of the thermal elements and transfers it
through natural circulation to the intermediate heat exchanger of circuits II–III; the coolant is speciallytreated water, which also acts as part of the steel-water radiation shield;- Circuit III is designed as a thermo-siphon with water or low-boiling coolant. Circuit III transfers heatthrough natural circulation to the heat exchanger of the heat supply circuit, the coolant being ethanol;- Circuit IV transfers heat from the heat exchanger of circuits III–IV to the consumers using forced circulation;
the circuit IV coolant is A-60 antifreeze.

5. Safety Features
The reactor is installed in a caisson forming a heat-insulating gas cavity in the strengthened area of the reactor
vessel and a caisson space above the reactor cover to house control and protection system (CPS) drives and toprevent radioactive substances from escaping into the surrounding space in case of a circuit I break. The
localizing safety systems provide defence in depth and secure the plant safety based on inherent safety featuresand predominantly passive phenomena; they require no human intervention or external power sources. The

316

safety barriers of the ELENA-NTEP are: (1) Fuel elements; (2) Leak-tight primary circuit; (3) Caisson; (4)Reactor vessel and the guard vessel designed to withstand the pressure arising within each of them at theirconsecutive failure; and (5) An embedded silo sealed with a protective plate. Special measures for theprotection of hot water consumers ensure that radioactivity is never released into the network circuit.

Engineered Safety System Approach and Configuration
ELENA systems are designed with inherent safety features to ensure it remains in a safe configuration under
any condition. The incorporation of the defence-in-depth approach based on six safety barriers prevents thedepressurization of the primary circuit from depressurization and secure activity confinement inside the reactor
during accidents. Though the use of a self-adjustable water-cooled reactor coupled with thermoelectric modeof heat conversion and natural circulation of coolant makes it possible to exclude movable elements from thetechnological circuit of a NPP and to secure a lifetime unattended operation without on-site refuelling. Safety
support systems create the conditions required for normal functioning of the safety systems; they include power
supply systems and a heat removal system that transmits heat to consumers. The active components of theprotection system are scram actuators for the six compensation groups of control rods.

Decay Heat Removal System
The low specific thermal power of the ELENA reactor enables easy removal of residual heat after reactor
shutdown. Residual heat is damped naturally to the compartment and the fuel elements are not super-heatedduring this process.

Emergency Core Cooling System
The control safety system (CSS) consists of a control safety system for emergency shutdown and a system to
input process and transmit safety-related plant information. During normal operation the emergency shutdown
CSS is permanently awaiting a scram actuation request; it also periodically provides information on the stateof the plant.

(d) Containment System

The reactor is installed in a caisson forming a heat-insulating gas cavity in the strengthened area of thereactor vessel and a caisson space above the reactor cover to house control and protection system (CPS)
drives, and to prevent radioactive substances from escaping into the surrounding space in case of a circuit I
break. In turn, the caisson is encircled by the external containment, which is the next barrier to the spread ofradioactivity; water that fills the containment volume is circuit II coolant and acts as a biological shielding
for the reactor. The external containment forms the cylindrical geometry of the plant with a height of 13 m
and a diameter of 4.45 m.

6. Plant Safety and Operational Performances
The ELENA reactor does not require an operator during nominal power operation of the plant. Operators are
required for assembly, startup and beginning of nominal operation. The reactor is designed to operate in a baseload mode. The reactor installation is based on passive principles of heat removal (natural convection in all
circuits, except for heat transport to the consumers) in normal operation and in shutdown conditions. A
decrease in heat or consumer power is automatically compensated through the discharge of excess heat to theatmosphere via a dry cooling tower, with no changes in the electric power. There are no valves or mechanicalparts which require maintenance over the lifetime of the plant.
Once operational the ELENA reactor depends upon natural processes to maintain the reactor power without
the actuation of control rods. The control and safety systems, including the control rods, control rod drive
mechanisms and sensors are used only for the reactor startup, or for the times that the reactor is scrammed.
Startup is done by an on-site operator who can leave the site once steady-state power has been obtained. The
reactor startup is done by measuring the neutron flux and calculating the reactor period. The reactor outlet
temperature and pressure in the coolant loop is monitored, but do not provide feedback through the control
loop during start-up. To begin the operation, the poison rods are pulled completely from the core, and are neverinserted during nominal operation. To start up and reliably shut down the reactor in any situation, a grid is
included that compensates the excessive reactivity. The compensation grid consists of six groups of the boron
carbide absorber rods in stainless steel claddings of 1.45 cm external diameter. Each group (34 rods) has anindividual drive.

7. Instrumentation and Control Systems
The instrumentation and control (I&C) system of the ELENA reactor is specially arranged to register parameter
deviations at early stages of the accidental conditions to predict their further progression.

8. Plant Layout Arrangement
The reactor system can be broken into two parts for shipment. It is possible to fuel the system on-site, thereby
eliminating problems associated with shipping a fuelled reactor. The plant includes instrumentation and control
systems; a system for heat removal to consumers; an auxiliary power supply system; and a radiation monitoring
system, including process radiation monitoring, dosimetric monitoring, and environmental monitoring.
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Reactor Building
The plant building has a cylindrical shape and is embedded in the ground for the entire reactor installation
height with a foundation plate elevation of –19.2 m. The elevation of +0.0 has a domed ceiling. The
underground portion of the structure, the walls and the overlaps are monolithic reinforced concrete. The vessel
head of the system is removable. The plant incorporates a physical protection system, has a fence and isequipped with external lighting.

Control Building
The plant has a main control and monitoring room accommodating the start-up and instrumentation and control
equipment, as well as the equipment necessary to prepare information to be transmitted to a monitoring centre.

Balance of Plant
i. Turbine Generator Building
A TEG is used as a heat exchanger between circuits I and II; it is based on semiconductor thermo-elements
enabling the generation of 68 kW of power in the reactor nominal operating mode simultaneously with heattransfer to circuit II. This power is used for plant auxiliary needs; it could also be supplied to a small town
without district power supply, partially replacing a diesel power plant. The TEG consists of eight identical
thermoelectric units (TEU). Each of them includes 36 thermoelectric modules equipped with thermoelectric
packs of bismuth tellurides with electronic and hole conduction.
ii. Electric Power Systems
The ELENA-NTEP CSS has three independent power supply systems, consisting of two (2) TEG sections, adiesel generator, and a storage battery. The electric power output can be controlled either by the use of shut
resisters or by short circuiting the TEs. The TE power conversion system has a low electrical conversion
efficiency, and the waste heat is used for district heating.

9. Design and Licensing Status
The assembly drawings of the ELENA have been completed and are ready for fabrication and testing of thesystem.

10. Fuel Cycle Approach
The factory-fabricated reactor vessel is delivered to the site loaded with fresh fuel. This initial load is designed
to provide the whole NPP lifetime without refuelling.

11. Waste Management and Disposal Plan
The waste management is not required during the ELENA-NTEP lifetime due to the safety barriers and noneed for maintenance. At its lifetime end, the reactor vessel is removed with the spent fuel in a shipping cask.
Liquid and solid radioactive waste is also disposed using special equipment. The site is either provided with a
new ELENA-NTEP or proceeds to "greenfield" status.

12. Development Milestones
Not determined.

318

UNITHERM (NIKIET, Russian Federation)
All content included in this section has been provided by, and is reproduced with permission of NIKIET, Russian Federation.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type

NIKIET, Russian Federation

Coolant/moderator

High purity water

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

30 / 6.6

NSSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array

16.5

Number of fuel assemblies in
the core
Fuel enrichment (%)

PWR

Natural circulation

249 / 330
UO2 particles in a metallic silumin or
zirconium matrix, metal-ceramic/ 54-55
265
19.75

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)

1.15

Reactivity control mechanism

Soluble boron and control rod insertion

Approach to safety systems

Hybrid (passive + active) system

Design life (years)

30

Plant footprint (m2)

~10 000

RPV height/diameter (m)9.8 / 2.9

RPV weight (metric ton)32

Seismic Design (SSE)

VIII-IX-MSK 64

Fuel cycle requirements /ApproachDistinguishing features

Traditional

Design status

200

Autonomous passive reactor decay
removal system; guard vessel; ironwater biological shielding; and thebiological shielding tanks
Conceptual design

1. Introduction
The UNITHERM is a small transportable nuclear power plant (NPP) with a capacity of 30 MW(t) and a rated
electrical output of 6.6 MW developed based upon NIKIET’s experience in designing marine nuclear
installations. The UNITHERM reactor is intended for electricity supply to urban areas and industrial
enterprises in remote regions. UNITHERM adopts a natural circulated primary cooling system and is intended
for minimal operational staffing with an option for unattended operation and a centralized regional support
facilities monitoring. The UNITHERM design adopts proven technology and operational experience of the
WWER type reactors. The design aims for fabrication, assembly and commissioning of the NPP modules tobe carried out at factory. The UNITHERM reactor is designed to operate for 20-25 years without refuelling as
both a land-based and barge mounted NPP. NPP with UNITHERM may consist of a number of units dependingon the purpose and demand of costumers need.

2. Target Application
The UNITHERM NPP can be used as a source of energy for the generation of electricity, district heating,seawater desalination and process steam production. In general, the configuration and design of the
UNITHERM is sufficiently flexible to be adjusted or modified for different target functions and user
requirements, without compromising the underlying principles of the design.
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3. Design Philosophy
NPPs with the UNITHERM reactor are designed for siting in remote regions with less developed infrastructure
and where qualified staff for plant operation may not be available. The reactor core life is expected to be equal
to the plant lifetime with an estimated time of 20-25 years. The refuelling of the core will not be required
during the plant service life.

4. Main Design Features
Nuclear Steam Supply System
Primary circuit system is intended for heat removal from the reactor core and heat transfer to the intermediate
circuit fluid inside the intermediate heat exchanger. The system consists of a main circulation train and apressurizing system. The natural circulation of primary coolant takes place in the primary circuit.
The intermediate circuit system is intended for heat transfer from the intermediate circuit coolant to thesecondary coolant (consumer’s circuit) inside a steam generator (SG). This system provides an additional
localizing safety barrier to protect the heat consumers against the ionizing radiation from radionuclides
generated by primary coolant activation, from structural material corrosion products dissolved in the primarycoolant as well as fission products entering the primary circuit in case of fuel cladding failure. Primary coolant
circulates by means of natural convection.
The secondary system (consumer’s circuit) is intended to generate a superheated steam from the feedwater(supplied by NPP turbo generator pumps) by means of heat transfer from the intermediate circuit coolant inside
the steam generator. Secondary coolant circulates by means of natural convection.

Reactor Core
The reactor core consists of 265 fuel assemblies installed in the plates of the removable reactor screen at the
points of a regular hexagonal lattice. The UNITHERM fuel element is designed as a cylindrical rod with four
spacing ribs on its outer surface. The fuel is in the form of tiny blocks of UO2 grains coated with zirconium
and dispersed in a zirconium matrix. The gap between the fuel-containing matrix and the cladding is filledwith silumin. A fuel element of such design has a high uranium content and radiation resistance. These features,
taken together, make it possible to operate such fuel elements during the whole specified core lifetime. A
specific feature of the UNITHERM fuel cycle is the long and uninterrupted irradiation of fuel inside the reactor
core throughout the reactor lifetime, without core refuelling. The metal ceramic (CERMET) fuel chosen for
the UNITHERM is composed of UO2 particles in a metallic (silumin or zirconium) matrix. This design ischaracterized by a high volume ratio of nuclear fuel; the use of the metallic matrix ensures minimum swellingand high thermal conductivity. Optimally shaped cladding is formed when the cladding is filled with the matrix
composition.

Reactivity Control
The control element drive mechanisms (CEDMs) are designed to provide secure insertion of rods in the coreby gravity for reactivity control. Locking devices are installed in the CEDM to avoid unauthorized withdrawalof control rods. Burnable absorbers are used to compensate the decrease of reactivity due to fuel burn-up,
temperature effect and by motion of the reactivity control rods during periodic maintenance.

Reactor Pressure Vessel and Internals
UNITHERM is an integral type reactor with nuclear steam supply system (NSSS) equipment installed inside
the reactor pressure vessel (RPV).

Reactor Coolant System
The UNITHERM primary cooling mechanism under normal operating condition and shutdown condition is by320

natural circulation of coolant. The heat energy released from the reactor core is transferred to the intermediate
circuit coolant, which moves upward to flow outside the tubes of the helically coiled once-through steam
generator (SG).

Steam Generator
The reactor employs a helically coiled once-through SG. Heat transfer from the reactor core to the intermediate
circuit coolant occurs in the built-in once-through intermediate heat exchanger and heat transfer from theintermediate coolant to the consumer’s circuit coolant – inside the SG. Both heat exchangers are made fromtitanium alloy. The intermediate heat exchanger has a structure of coil bundle consisting of 80 separate
subsections that are united in 8 independent sections. Their supply and discharge tubes are connected to 8
pressure vessel steam generating modules installed on the reactor cover.

Pressurizer
Pressurizer of UNITHERM is a built-in structure of the upper plenum of the RPV.

5. Safety Features
The UNITHERM safety philosophy is to assure that the radiation impact on personnel, population and theenvironment under normal and design basis accidents is well below the limits prescribed by the current
regulations. The UNITHERM design makes use of passive systems and devices based on natural processes
without external energy supply. The design inherently eliminates potentially hazardous activities related to thecore refuelling, as the reactor core refuelling will not be required in the plant service life. This further simplifies
the operating technologies and enhances the proliferation resistance.

Engineered Safety System Approach and Configuration
The UNITHERM safety systems are based upon redundancy, diversity and the maximum use of the fail-safe
systems. The UNITHERM employs passive safety systems and devices which do not require actuation (suchas containment, independent heat removal system, etc.) or can be passively actuated (such as primary circuit
systems and containment depressurization system). The reliability and safety of the UNITHERM reactor issignificantly improved due to the elimination of the shut-off and isolation valves from the reactor pipelines,
except for the user circuit, i.e., all systems are in continuous operation. The component cooling circuit ispassively operated and continuous removal of heat from the reactor components enclosed in the containmentis achieved efficiently. The structures of the UNITHERM NPP are designed to protect the reactor from extreme
external events such as hurricanes, tsunami, aircraft impacts, etc. The reactor can be automatically shut down
and brought to a safe state without exceeding the design limit. The UNITHERM also incorporates several
design features and measures for protection from human errors and mitigation of the consequences of humanerrors or acts of malevolent.

Emergency Core Cooling System
An independent passive heat removal system is adopted which acts as a cooldown system in emergency
shutdown of the reactor. During a postulated loss of coolant accident (LOCA) scenario, some primary coolant
and steam-gas mixture from the pressurizer are discharged to the containment. The emergency core protection
system is activated in response to the signals from pressure transducers. Coolant leakage continues until thepressure values in the reactor and containment are equalized. The remaining coolant inventory in the reactor
is sufficient to maintain circulation in the primary coolant circuit. The reactor is passively cooled via theintermediate circuit and the independent heat removal circuit, whereas the containment heat is removed by thecomponent cooling system. Additionally, an active user circuit with feedwater supplied to the SG and steamwater mixture maybe utilized to increase the cooling rate. The iron-water biological shielding acts as a systemof bubble tanks for cooling water storage. The shielding removes heat from the RPV, preventing a core melt
in a postulated beyond design basis accident with reactor core voiding.

Containment System
The integral reactor for land-based deployment is placed inside theleak-tight containment, which is located within the concrete shockresistant structure together with the biological shielding and reactor
unit components. This structure enhances physical protection of thereactor unit from external impacts such as airplane crash, hurricane,tsunami, unauthorized access, etc.
(1) Iron-water shielding tank;(2) Containment;(3) Shock-proof casing;(4) Cooldown system heat exchanger;(5) Safeguard vessel(6) The reactor
The containment system is capable of maintaining the primarycoolant circulation as well as provides reactor cooldown and321

retention of radioactive products under the loss of primary circuit leak tightness. Passive safety systems for theremoval of heat from the containment and biological shielding tanks are employed.

6. Plant Safety and Operational Performances
Electrical output of the NPP with UNITERM-30 reactor equals to N(e) – 6.6 МW(e). Electrical voltage
provided to the user grid – alternate 3-phase 10.5 кV ± 10 %, frequency 50 ± 1 Hz. Basic regime of NPP
operation lies within the power range from 20 to 100 % N(e) providing a daily and annual load following. The
speed of power augment and drop – 0.1 % от N (e)/seс. Upon a customer request there may be foreseen
additional provision of thermal power. Maximum rate of it could be up to 28 Gcal/hour.

7. Instrumentation and Control Systems
Аutomatic System of control for technological processes of the NPP allows for:- Safe operation of the NPP and electrical generation; protection from the violations of safe operation limits
and conditions; prevention of accidents; mitigation of accident consequences; bringing the NPP back to thecontrolled and safe condition during accidents and after them.
- The Automatic Control System consists of functionally completed systems developed on the basis ofprogrammatic-technical systems and instruments that were trialed in the NPP conditions or other analogous
objects.
Technical appliances for the ACS are manufactured at the enterprises according to approbated technology andmethods of testing and control while strictly observing the requirements of quality control.

8. Plant Layout Arrangement
Reactor Building
The NPP site is limited by perimeter of the protected zone that does not exceed a square of 2 hectares.
The site hosts reactor building for housing reactor(s) which possesses special transport locks for delivery of
the reactor plant for mounting and other equipment necessary during outages and removal of the reactor
facility; building to house turbine-generator(s) and some other auxiliary buildings. Turbine-generator assemblyfor UNITHERM NPP depends on the plant capacity and operation mode requested by its users. The turbine
operates using dry saturated steam in the mode of steam outlet backpressure. With consideration of thecontinuous transfer of 5 % heat to the independent heat removal system, the total efficiency in this case is
expected to be ~74 %. High efficiency is achieved from the utilization of low-parameter heat at the turbine
exhaust. An electric generator with an output of 6.6 MW(e) in combination with a single-phase intermediate
circuit allows to obtain a superheated steam temperature of 285°С under 1.35 MPa.

9. Design and Licensing Status
Based on the experience of NIKIET and other Russian institutions and enterprises in the development ofmarine nuclear installations, the UNITHERM NPP may require no major research and technology
development activities for its deployment. Once an agreement with the user is reached and the technicalassignment approved, it is estimated that 5 years will be required to finalize design development, licensing,construction and commissioning of the UNITHERM NPP, provided there are no financial or organizational
constraints. The detailed design stage would include qualification of the core, heat exchangers, CEDMs andother components.

10. Fuel Cycle Approach
The duration of the campaign reactor core is 15 years.

11. Waste Management and Disposal Plan
Fuel handling is based on the traditional scheme implemented for the marine-based prototype reactor. Fuelprocessing and disposal will be performed at a specialized enterprise.

12. Development Milestones
1994
2012
2015

322

The NPP design on the basis of the UNITHERM concept has become the laureate of thecompetition on SMR designs established by the Russian Nuclear Society
Technical proposal on the UNITHERM reactor facility (WDR stage)
Technical proposal for a SMR plant based on the UNITHERM reactor

Advanced Micro Reactor – AMR(STL Nuclear (Pty) Ltd, South Africa)
All content included in this section has been provided by, and is reproduced with permission of STL Nuclear (Pty) Ltd, South Africa.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
STL Nuclear (Pty) Ltd., South Africa
country of origin
Reactor type
Coolant/moderator
Thermal/electrical capacity,
MW(t)/MW(e)HTR, Advanced Prismatic Type
Helium, graphite
10 / 3 for single module plantPrimary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
4/1

Core inlet/outlet coolant
temperature (oC)450 / 750

Fuel type/assembly array
Number of fuel assemblies
in the core
Fuel enrichment (%)
Refuelling cycle (months)
Core Discharge Burnup(GWd/ton)

TRISO particles/ LBE eutectic/ SiC tubes
420

Reactivity control
mechanismApproach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)

Control and shutdown rods in core,
and the reflector
Passive / inherent decay heat removal
40
2 500
5.96 / 2.78
115
0.3g (generic site)
0.5g under consideration
LEU UO2

Fuel cycle
requirements/approachDistinguishing features
Design status

10 to 209680-90

Additional SiC barrier; Lead Bismuth
Eutectic (LBE) heat transfer medium
Pre-conceptual design

1. Introduction
STL Nuclear (Pty) Ltd., the University of Pretoria, the North-West University in conjunction with the South
African Nuclear Energy Corporation (NECSA) are developing a 10 MW(t) Small Modular Reactor (SMR)
called the Advanced Micro Reactor (AMR). This reactor falls in the category of the High-Temperature, Gas
Cooled Reactors (HTGRs). The AMR uses helium as the coolant and is graphite moderated. It uses graphite
hexagonal blocks as the moderator and these blocks are arranged to form a cylindrical configuration. The
individual SiC fuel assemblies contain TRISO coated particles with either uranium dioxide (UO2) or uranium
oxycarbide (UCO) ceramic fuel kernels of between 10 wt% to 19.9 wt% enriched uranium. The voids
between the coated particles are filled with a Lead Bismuth Eutectic (LBE) alloy to provide good heattransfer from the fuel particles to the fuel assembly wall.

2. Target Application
The AMR can supply electric power to any distribution grid and to standalone or isolated electricity users. It
can be deployed as single modules or multi-module plants as well as for process heat applications.

3. Design Philosophy
AMR design utilizes proven HTR technologies albeit in a different configuration The reactor is designed tohave excess reactivity to operate for several years before refuelling is required. The design is factoryassembled to enable road transportation.

323

4. Main Design Features
Power Conversion
A Heat Pipe Heat Exchanger (HPHE) is used to transfer heat from the reactor core to the secondary power
conversion loop containing air. The LBE alloy is used to absorb heat from the primary helium coolant bundle(located in the centre) and surrounded by a flow-directing shroud. This heat is then exchanged by means of
natural convection with the secondary power loop (in the form of a multiplicity of U-tubes) located in theannulus between the flow-directing shroud and the inner wall of the outer shell of the heat exchanger. The
LBE returns to the central hot helium bundle at the bottom through several portals in the flow-directing
shroud. This flow of LBE in this design can be described as "toroidal natural convection flow".

Reactor Core
The AMR is a high temperature helium-gas cooled
reactor with a power of 10 MW(t). The helium
coolant is circulated through the reactor core by anelectric blower located within the pressure boundaryof the heat exchanger. As is the custom for thermal
neutron-spectrum high temperature reactors, graphite
is used as moderator, which in this design, is in theform of hexagonal graphite blocks packed to form anapproximate cylindrical configuration with adiameter and height (including radial reflectors aswell as top, bottom and) respectively of ~2.2 m(~1.72 m active diameter), and 2.4 m (active length
2.2 m) with a volume of 5.103 m3 which results in acore power density of ~2 MW/m3. The fuelassemblies are silicon carbide tubes that contain thefuel in the form of low-enriched uranium (LEU)
dioxide (UO2) or uranium oxycarbide (UCO)
TRIstructural-ISOtropic (TRISO) coated particles,
immersed in an LBE (45% Pb and 55% Bi) alloy.

Heat Pipe Heat Exchanger (HPHE) coupled to the AMR

Reactor Core

Fuel Characteristics
In typical coated particle. The SiC is the main layer for the retention of fission products. The AMR is also
not only limited to kernel and coated particle diameters of 425 microns (μm) and 855 μm respectively but
can also utilize kernel diameters of 500 μm and coated particle diameters of 920 μm. The AMR is not
restricted to using only high assay low enriched uranium (HALEU) it can utilize enrichments of ~10 wt% to
15 wt%. Approximately ~463 775 coated particles will be contained in a typical fuel assembly.

Fuel Handling System
The AMR is designed to operate between 8 to 10 years without the need for refuelling. Burnable absorbers
are used to supress the flux at the beginning of life and extend the life of the reactor as they deplete. The
reactor at time of refueling, will be transported back under IAEA seals to the NECSA site where the fuel will
be unloaded under continuous IAEA containment and surveillance measures. The spent fuel will be removed
and stored during the initial post-operational cooling period on-site in shielded vessels under IAEA seals.
After this initial cooldown period and depending on the spent fuel storage strategy in force at the time, it willbe transported to its final storage location. A new core will be loaded in the reactor vessel using the sameprocess as described, and the newly loaded vessel will be transported its operational site.

Reactivity Control
As a defence-in-depth, two diverse shutdown systems, each capable of shutting the reactor down with fresh
fuel when the reactor is cold and with one member (of highest reactivity worth) stuck in the full extracted
position. The shutdown margin is 8%. Thirteen neutron absorber rods are provided in graphite sleeves; 7
within the core and 6 within the side graphite side reflector blocks. There are also 6 shutdown rods locatedwithin the core. The absorber rods can be operated independently as a group or as sub-groups, as required by
the reactor operating control system. A control rod consists of several rod absorber material segments,
pinned together to form articulating joints. The segments consist of sintered B4C absorber material,sandwiched between an inner and an outer tube segment. The inner tube segment allows cooling helium gas
to flow form the top down in the circular channels. If all control and shutdown systems are accidentallywithdrawn, it will not lead to fuel damage. This is normally defined as a "start-up accident".

Reactor Pressure Vessel and Internals
The Reactor Pressure Vessel (RPV) is constructed in compliance with the ASME III subsection NB code. It
comprises two main components reactor of vessel body and vessel head which is bolted to the vessel body.
The reactor vessel body consists of several forged ring-components circumferentially welded together.
324

The core structures consist of the metallic parts and the graphite structures. The function of these internal
structures is to provide stable core geometry, neutron reflection, cold and hot gas channelling, fuel element
flow, shielding, thermal insulation and support of the control and shutdown systems guide tubes and theneutron source. The functional design of the structural core internals is such that they can withstand thesteady state and transient loadings during normal operation, anticipated operational occurrences and designbasis accidents. The loads borne by the ceramic internals are transferred to the stainless-steel core barrel andthen to the reactor pressure vessel through metallic components such as the lower support structure and thecore barrel axial and radial supports. All areas of the core internals are designed for the service life of thereactor. Access for ceramic structure inspections can be done through the fuel loading channel and thereflector rod holes.

Reactor Coolant System
The helium coolant enters the reactor vessel at 450°C through the annulus of the co-axial duct. The helium
then flows upwards in the helium risers located in the outer graphite reflector. Helium leak flow also enters
the annular space between the Core Barrel (CB) and the inside of the Reactor Pressure Vessel (RPV). The
helium flow enters the top of the reactor core where it is evenly directed to the 420 borings containing thefuel assemblies as well as between the annuli of the control rod guide-tubes. The helium leak flow also enters
the top and bottom RPV domes. The helium from the bottom dome re-joins the leak annulus between the CB
and RPV, while the helium entering the upper dome flows past the control rod guides and metallic
components and connects to the upper gas plenum and is routed downwards to join the major coolant flow
past the fuel assemblies. Some helium from the upper dome is also forced into coolant holes within thecontrol rod guide sleeves which flows in the inner anulus of the control rods, this then exits the sleeve and rejoins the major downward flow in the core. The helium then flows downward through the annulus in theborings past the fuel assemblies to remove heat and exits the core at 750°C. It is then collected in a lower
core hot gas plenum that is part of the lower core support structures and flows back through a hot duct(connected to the hot gas plenum) to the HPHE.

5. Safety Features
Engineered Safety System Approach and Configuration
The plant is designed to perform its safety functions without reliance on the automated plant control system,or the operator. The plant has no engineered safety systems in terms of active human or machine intervention
to assure nuclear safety.
Nuclear stability: If all control and shutdown systems are accidentally withdrawn, it will not lead to fueldamage or a radionuclide release. There is no requirement for active safety systems or operator action toprevent fuel damage. This is achieved with a relatively large negative temperature coefficient of reactivity
over the entire operational range, a low core power density, a core geometry that will ensure passive decay
heat removal and the radionuclide retention capacity of the TRISO particle fuel as well as the SiC structureof the fuel assembly. Xenon oscillations is damped due to the H/D ratio of the core of less than 3 which isnormally used as guideline for inherent stability.
Thermal stability: The low power density is ensured in the core design as well as a high thermal capacity andheight to diameter ratio (H/D) greater than 0.97 to ensure that the decay heat removal can solely be achieved
through conduction, natural convection, and radiation through the reactor structures. It was determined that
increasing the height/diameter ratio beyond 0.97 in a trade-off between neutron losses versus the advantage
of gaining passive decay heat removal via the walls of a steel pressure vessel in the event of a DLOFC.
Mechanical stability: The design also ensures that the materials of construction remain below the structuraldesign limits and the maximum fuel temperatures in an accident condition remain below the set fuel damage
limits

Decay Heat Removal System / Reactor Cooling Philosophy
The Reactor Cavity Cooling System (RCCS) removes heat radiated from the reactor towards the reactor
cavity walls. It consists of welded membrane tubes arranged side-by-side on the inside of in the reactor
cavity wall. Water is circulated through the tubes to form a cold wall. The RCCS is a passive system andconsists of independent cooling trains and is designed for all postulated design basis conditions.

Spent Fuel Cooling Safety Approach / System
The spent fuel will be stored in special tanks of which the height/diameter is in excess of 4. Thermosiphons(heat pipes) are attached to the outside walls of these tanks which removes heat to the outside of the spent
fuel area. The condenser ends of these heat pipes are then fitted with fins to dissipate the heat to atmospherein an entirely passive way. There is also the possibility to use Stirling engine/generators on the condenser
ends to generate electricity for charging batteries and in so doing provide a measure of energy conservation.

Containment System
The reactor has two barriers of SiC, that being the SiC layer surrounding each fuel kernel as well as the SiC
fuel tube of the fuel assembly. This introduces an additional barrier which will further reduce the probabilityof any radioactivity being released into the environment. In addition to the two abovementioned barriers, the325pressure boundary serves as a third barrier against release of radioactivity, and then as a final barrier thereactor building can also be accredited as a final enclosure.

Chemical Control
Chemical stability: The design of the core and its coolant routing is such that in an event that could allow air
to leak into the pressure boundary, there is no possibility that a sustained corrosion of core components by air
can take place. The reactor also does away with the possibility of a water or steam ingress scenario as thehelium coolant will transfer heat to a HPHE which is a single-phase natural convection heat pipe heatexchanger using LBE as working fluid. This heat exchanger is then coupled to a Brayton power conversion
cycle. The use of a HPHE also introduces another important safety feature by eliminating the possibility oftritium, produced in the primary helium cooling circuit to contaminate the air in the secondary circuit bydiffusing through a single tube wall.

6. Plant Safety and Operational Performances
The thermal-hydraulic calculations show the maximum fuel temperatures for the calculations was 1101.5°C
and the maximum central fuel temperatures determined using a validated computer code being 1084°C (adifference of only ~17.5°C) which is 1.6%. LBE is used as filler material in the fuel assembly due to the factthat it reduces the central fuel temperature by ~130°C compared to a fuel assembly only containing helium as
the filler material. The fuel will remain below the normal operational temperature guideline chosen of
1130°C and in the case of a DLOFC event will remain below the Germany set temperature limit of 1600°C.

7. Instrumentation and Control System
The Automation System (ATS) comprises the group of safety and non-safety I&C systems that provide
automated protection, control, and human-system interfaces. Three specific systems in the AMR that define
I&C are plant control, data and instrumentation system, investment protection system and protection system.

8. Plant Layout Arrangement
The reactor building contains the safety equipment that provides the necessary functions for the safe
shutdown of the reactor under all design basis conditions. The reactor building (RB) is partially submerged
below ground level such that the reactor and heat exchanger cavities are completely protected against
postulated external threats. The reactor building, electrical building and auxiliary buildings are connected bymeans of underground tunnels, providing protection for interlinked services. The RB is seismically designed
to withstand a design basis earthquake (DBE) and is the only safety related building structure of the AMR.

9. Testing Conducted for Design Verification and Validation
Experimental activities are being conducted to determine heat transfer enhancement of the Lead Bismuth
Eutectic (LBE) in SiC fuel tubes. It is anticipated that the core neutronics will be verified in a future criticalfacility at the NECSA site. The Helium Test Facility on the NECSA site will be used to test the operational
performance of key components under non-nuclear operational conditions of temperature, flow and pressure.

10. Design and Licensing Status
The AMR is at pre-conceptual phase. The neutronics, thermo-hydraulics and heat transfer analyses are being
conducted to verify the safety analysis. Mechanical design and material selection are also underway.

11. Fuel Cycle Approach
The AMR is designed for UO2 and UCO. However, it is not limited to the use these types of fuel. The AMR
while maintaining safety characteristics, can use alternate fuels without modifications. Advanced fuel cycles
for later investigation may range from a (Th, U)O2 using both LEU and HEU, to a UC2 fuel cycle.

12. Waste Management and Disposal Plan
Disposal of spent fuel elements from the AMR is performed in the following sequence: (i) Direct transfer ofspent fuel elements into a flask inside the cast iron high energy spent fuel casks (Hi-cask); (ii) Immediately
after filling the H-cask they are sealed and transferred to the spent fuel cool-down facility; (iii) Once cooled
down, the flask filled with fuel is transferred from the Hi-cask to a low energy spent fuel concrete cask
(Low-Cask); iv) The Lo-Cask is transported to the low energy on-site interim storage facility; v) For offsite
transport the flask is transferred to a shipping/transport cask for shipping to an ultimate repository.

13. Development Milestones
2020
2021
2022

326

Project Started
Pre-conceptual design
To begin with conceptual design

LFR-TL-30(newcleo Ltd, United Kingdom)
All content included in this section has been provided by, and is reproduced with permission of, newcleo Ltd, United Kingdom.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
newcleo Ltd, United Kingdomof origin
Formerly developed by Hydromine
S.àr.l, Luxembourg
Reactor type
Liquid metal cooled fast reactor
Coolant/moderator
Lead / none
Thermal/electrical capacity,
90 / 30
MW(t)/MW(e)Primary circulation
NSSS operating pressure(primary/secondary), MPa

Forced circulation
0.1 / 14 (absolute)Core inlet/outlet coolant
temperature (oC)420 / 530

Fuel type/assembly array
Number of fuel assemblies in
the core
Fuel enrichment (%)
Refuelling cycle (years)
Core discharge average burnup(GWd/ton)

UO2 / hexagonal fuel assembly
37

Reactivity control mechanism

6 x 30° circular sectors made ofabsorber material

Approach to safety systems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)
Fuel cycle
requirements/approach

Active + Passive
30
N/A
5 / 3.7~ 35
0.5g
Wastes burned in another newcleo
model- Active + passive safety;- No intermediate loops; Simple,- compact primary system andcompact reactor building.
Conceptual design

Distinguishing features

Design status13.0 % inner fuel, 19.75% outer fuelOne batch, each 15 years
49

1. Introduction
The LFR-TL-30 is an innovative micro SMR (Small Modular Reactor), cooled by molten lead, conceived tocomply with the GEN IV reactor criteria of safety, sustainability, proliferation resistance, and economics
despite the unfavourable small-scale effect. LFR stands for lead-cooled fast reactor, TL stands forTransportable Long-lived and 30 is the net electricity in MW. The LFR-TL-30 is an integral pool-type fast
reactor with all the primary components installed inside the reactor vessel (RV). Main primary components are
the innovative core, the innovative Spiral-Tube Steam Generator (STSG), the mechanical pump (MP),reversed-flag type control rods, decay heat removal by dip coolers and a safety vessel air cooling system. The
reactor design is derived from the LFR-AS-200 with the objective to verify up to which extent it is possible toapply the innovations embodied in the larger LFR-AS-200 to design a micro reactor with similar level ofcompactness. Several options regarding the fuel type, the power level and the thermal cycle, which are different
in relation to LFR-AS-200, have been investigated. An important cost parameter to be considered when
designing a micro reactor is the plant cost per unit power ($/W). It is possible, namely, to reduce the reactor
size while reducing power, but the plant $/W ratio is likely to become prohibitively high, owing to the cost of
the fuel handling machines and the buildings and facilities for storage of the fresh and spent fuel assemblies,
which is relatively independent from the reactor power and hence increases the $/W ratio. It should beconsidered, too, that it is not wise, for risks of proliferation, to provide the predictably numerous micro reactor
plants with these fuel storage facilities. A measure to overcome this proliferation and cost issue is to design327

micro reactors capable to be transported, complete of the core, to centralised facilities for fuel handling andmaintenance of main components. For this design approach to become viable, the micro reactor shall be
provided with a long-life core and be capable of transport in upright position, in order not to affect its
mechanical and thermal-hydraulic configuration while traveling. The LFR-TL-30 can be designed to comply
with both features of long- life core and transportability in upright position. Long-life cores are possible owing
to the high breeding capability of the fast reactor; transportability, that is bound to the compact reactor assembly,
in particular to the short outline height, is a merit of the very compact pump-spiral-tube steam-generator(pump-STSG) assembly previously conceived for the LFR-AS-200. No on-site refuelling being requested, it
is possible to install only one pump-STSG assembly centreline above the core. This is the main characteristicof the LFR-TL-30 reactor, which allows to balance the unfavourable scale effect with respect to the LFR-AS200 and to approach the unprecedented title of merit of a specific volume of the primary system of about 1 m3/MW(e) which is an outstanding feature of the LFR-AS-200.
LFR-TL-30 (and LFR-AS-200) will profit of the development of the LFR-AS-30 research reactor, designed
specifically by newcleo to support the design and qualification processes of the innovative solutions proposed.

2. Target Application
Target applications include: Sites without interconnected grids; Offshore oil platforms; Mines; Islands;
Naval propulsion (after appropriate modifications).
These sites require an “inherent safety” with the associated no need of the “emergency zone” and a particularsimplicity of operation for staffing reduction. The deployment of a micro reactor is facilitated by the uniquesafety features of the LFR, particularly for applications in oil platforms and/or naval propulsion: in exceptional
accidents, like ship collision and sinking, frozen lead will always maintain the confinement of the core withoutwater pollution and preserving potential for reactor recovery

3. Design Philosophy
The LFR-TL-30 is an integral pool-type fast reactor with all the primary components installed inside the reactor
vessel (RV). Main primary components are, the innovative STSG, the mechanical pump (MP), flag type control
rods, dip coolers of the decay heat removal (DHR) system. Thanks to the properties of lead, intermediate loops
can be eliminated with several special precautions to deterministically eliminate any risk of important primarysystem pressurization: among them water and steam collectors located outside of the RV, and a short STSG.
Reactor shutdown and DHR is performed by means of diversified and redundant systems which are passively
operated, and which can also be passively actuated when the primary system exceeds certain pre-set threshold
temperatures.

4. Main Design Features
Power Conversion
The heat is directly transferred from the primary lead to the water/steam system by means of a single STSG.
Superheated steam is produced and used in a turbine (or other steam consumers if needed).

Reactor Core
The core consists of 37 wrapped, hexagonal FAs, each containing 469 fuel pins laid out on a triangular pitch.
The core is removable for replacement by a fresh core in an off-site facility. The heads of the FAs areinterconnected by means of cams, which are integral part of each head. The result is a self-sustaining coreanchored to an internal structure (the inner vessel). The FA foot is free from mechanical supports (no core gridof classical design such as the Diagrid of SPX1) except for the radial contact with adjacent FAs, which createa packed bundle, contoured by the bottom end of the inner vessel.

Fuel Characteristics
Fuel considered is HALEU in oxide form, although alternative fuels, like metal, nitride or carbide fuel, canalso be considered. To comply with non-proliferation norms, the highest enrichment is kept below 20%(19.75%). The consequence is that for very small reactors like LFR-TL-30, the volume of the core is mainly
dictated by the need of a sufficient mass of fuel to reach criticality and ensure a reactivity margin to compensate
for the reactivity swing during burn up. The mass of required uranium is in the range of 8.6-9.6 t.

Reactivity Control
The reactivity control during normal operation is performed by actuation of ex-core absorbers which caneffectively compensate the criticality swing. The control devices are located in the space between the inner
vessel which supports and contain the core and the core itself.

Reactor Pressure Vessel and Internals
The RV is shaped as a cylindrical vessel with toro-spherical bottom head and flat roof. The free surface of leadis kept sufficiently below the roof to allow for a gentle thermal gradient between the vessel in contact withlead and the colder roof. The plenum above the free level is filled by argon as cover gas. The roof is made ofa circular thick plate with small-diameter penetrations for the dip coolers of the DHR and of the control andshutdown devices, and a large-diameter central penetration for the Spiral-Tube Steam Generator with the328integrated Primary Pump. The inner vessel is shaped as a double bottom shell and comprises a substantially
cylindrical lateral wall supported by the roof and upper bottom lower bottoms extending radially inwards fromthe lateral wall. The lower bottom has a central opening delimited by a peripheral edge for radial constraint of
the Fuel Assemblies (FAs) of the core. The upper bottom internally comprises a plurality of jaws for verticalsupport and elastic radial constraint of the FAs. The FAs are fastened together on the heads by latches (barspivoting into catch-slots) provided in their heads and are provided with a plurality of thermal expanders. At
normal operating condition, the front plates of all thermal expander are in touch with each other, but idle. On
trespassing a certain temperature limit chosen at the design phase, FAs are pushed radially outwards to reducecore reactivity and bringing core power into balance with the heat dissipation capacity of DHR systems.
All internals are hung to the reactor roof and have no connection with the reactor vessel. The MP, STSG, inner
vessel and core are co-axial with the reactor vessel.

Reactor Coolant System
The vertical axial-flow MP is installed on the RV centreline in the available space inside the tube bundle of
the STSG. The pump rests on and is connected to the upper support plate of the SG by means of a flange whichcloses the pump’s shaft penetration through the reactor roof and supports the variable-speed electric motor of
the pump. The pump is characterized by a short, large-diameter, tapered hollow shaft containing lead broughtin rotation by the shaft itself, in order to increase the mechanical inertia of the pump. There are no in-leadpump bearings. Primary lead circulates inside the core from the bottom to the top, then is conveyed by a funnelshaped structure to feed upward the MP and then the STSG, which is thereby fed from the bottom. Hot leadflows radially through the perforated inner shell of the Steam Generator and, once past the tube spirals,
continues flowing radially through the perforated upper portion of the inner vessel, and then downwards
between the inner wall of the RV and the outer wall of the lower part of the inner vessel to feed again the core.
The RV is always in contact with cold lead below the creep temperature of the steels.

Secondary System
There is no intermediate loop between primary lead and water/steam system. The elimination of the need for
an intermediate coolant system to isolate the primary coolant from the water and steam of the energy
conversion system represents a significant advantage and potential for plant simplification and improved
economic performance.

Steam Generator
The STSG is an innovative SG conceived for compactness and because it offers several advantages in termsof reactor cost, safety, reactor operability and simplicity of the lead flow path. The SG tube bundle is composedof a stack of spiral-wound tubes, arranged one above the other and equally spaced. The inlet and outlet endsof each tube are connected to the feedwater header and steam header, respectively, both arranged above thereactor roof to eliminate, in case of their failure, the risk of large water/steam release inside the reactor vessel.
The SG is thermally almost equivalent to a pure counter-current SG, because the feedwater in the tubes
circulates from the outer spiral to the inner spiral, while the primary coolant flows radially in opposite direction
from the inner shell to the outer shell. Because the flow path of the primary coolant inside the bundle is short,
its speed can be increased while keeping the pressure loss limited.

5. Safety Features
Engineered Safety System Approach and Configuration
In a reactor cooled by lead there is a large margin between the operating temperature and the safety limit andthe LFR-TL-30 exploits this margin for actuation of passive shutdown and passive decay heat removal systems,
which do not need power sources, operator intervention and logics. The shutdown system is realised by two
diversified systems. The first is an ex-core system envisaged to promptly shutdown the reactor in case of
emergency. It is formed by an absorber pin bundle kept together by two main grids located at both its ends;
the top grid is connected to the holding shaft, with the main function to keep the bundle in withdrawn position
and to insert it in case of SCRAM signal. The system indeed, operates in a traditional way, being driven fromgravity by means of a ballast. The absorber portion is also made of highly enriched B4C. The second redundantsafety shutdown system, unpublished, is passively actuated and passively operated. In case of unprotectedtransient with local clad failures, the steam-cladding accidental reaction and the resulting generation ofhydrogen and associated explosions, encountered with other liquid metal concepts, are excluded with leadcoolant. Finally, lead has also good retention capability of volatile fission products and in extreme conditions
the reactor can be directly cooled by jets of water, with the further advantage that frozen lead builds up its own
sarcophagus (barrier made of frozen lead) and definitively stop radionuclide dispersion.
The ultimate goal is the elimination of the need of an emergency preparedness zone.

Decay Heat Removal System / Reactor Cooling Philosophy
DHR is performed by means of two diverse (DHR1 and DHR2), redundant systems, each consisting of two
identical loops. One loop is adequate to remove the decay heat. The DHR1 system (not yet disclosed) removes
heat through the cold collector of the primary system. Each loop of the DHR2 system is equipped with spirals
of square-cross-sectional tubes wrapped around the reactor safety vessel for transfer to a water-steam system329

the heat transmitted by radiation from the reactor vessel to the safety vessel. The steam is passively condensed
in an air cooler which can be actively actuated or even passively actuated above a certain temperature threshold.

Spent Fuel Cooling Safety Approach / System
The plant is not provided with refuelling capabilities (flasks, pools). The reactor must be transported in uprightposition in a refuelling factory where these facilities are foreseen. Spent fuel pool are supposed to work in“dry” conditions.

Containment System
The reactor is provided with a concrete containment external-missile-proof. The dimension of the containment
is kept small by the very low potential energy stored in the coolant (which operates at atmospheric pressure)and the small inventory of water/steam of the secondary circuit. A safety vessel eliminates any loss of coolant
accident (LOCA) even in the event of a failure of the reactor vessel.

Chemical Control
An oxygen control system is implemented to avoid the formation of lead oxides and at the same time thepreserve the passivation of steel structures and components. Several systems, already successfully tested on
experimental loops, are under investigation for implementation in the research reactor LFR-AS-30 and then in
the industrial reactor LFR-TL-30.

6. Plant Safety and Operational Performances
Taking into account the envisaged uses of this type of reactors, a major aim of the design is to avoid the needof safety-grade diesel-generators and, for example, to use multiple smaller units on the same site, in order toimprove the electrical network availability, preferring the use of UPS (batteries) with an important capacity.
The expected refuelling cycle is about 10 to 15 years, without the need of important maintenance outages in
between. The system is conceived to operate mainly in reactor priority mode, but the potential for operation inload follow mode is under investigation.

7. Instrumentation and Control System
Core instrumentation (mainly thermocouples) will be used by the automated control and surveillance systemand by the protection one, with due attention to diversification at sensor and platform levels. Remotesurveillance and shutdown capabilities will be implemented.

8. Plant Layout Arrangement
The key for economics of the LFR-TL-30 is based on reactor compactness and associated compactness of thereactor building. The compactness of the reactor, the absence of intermediate loop, of the fuel handlingfacilities and spent fuel storage allows to drastically reduce the size of the reactor island civil structures (be iton land or on a ship).

9. Testing Conducted for Design Verification and Validation
The LFR-TL-30 is in the conceptual design stage. Most of design verification and validation will be made on
LFR-AS-30.

10. Design and Licensing Status
Licensing of the LFR-TL-30 has not started, yet. Licensing will take advantage of the return of experience ofthe design and operation of the LFR-AS-30 under fast development.

11. Fuel Cycle Approach
In a scenario of revived interest in nuclear energy, the LFR-TL-30 fuel cycle would naturally fit in an assumed
large nuclear energy use programme of reactors of various sizes and for various uses. In this scenario the spent
fuel could be reprocessed for the recovery of the important residual fraction of 235U to be reused in transportable
reactors, and of plutonium and possibly also of minor actinides as a fuel of central-station fast reactors.

12. Waste Management and Disposal Plan
After fuel reprocessing and Pu recovery (to be burnt for instance in the LFR-AS-200 always proposed by
newcleo for this purpose), wastes are meant to be disposed. At the time of refuelling, the components of the
primary system can be reused after inspection and requalification. Lead will be recycled in other reactors.

13. Development Milestones
2017

2019
2021

330

Completion of pre-conceptual desing (by Hydromine Nuclear Energy S.àR.L,
Luxembourg)
Conceptual design started
Hydromine Nuclear Energy S.à r.l. is incorporated in newcleo Ltd for a fast development
program. Design restarted after an idle phase.

U-Battery (Urenco, United Kingdom)
All content included in this section has been provided by, and is reproduced with permission of Urenco, United Kingdom.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, countryof origin
Reactor type
Coolant/moderator

Urenco, United Kingdom
High-temperature gas-cooled micro
nuclear reactor
Helium / Graphite

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation

10 / 4

Secondary cycle

Nitrogen secondary circuit (nowater)TRISO / Hexagonal

Fuel type/assembly array

Forced circulation helium

Number of fuel assemblies in
the core
Fuel enrichment (%)

12 x 5

Core Discharge Burnup(GWd/tonne)
Reactivity control mechanism~80 (average)< 20%

Design life (years)

Control rods, fixed burnable poisons,secondary shut-down absorber
spheres
5 EFPY core life, 30 year design life

RPV height/diameter (m)Approx. 5.5 / 2.2

Distinguishing features

Simplicity, established technology
basis, demonstrated fuel.
Multiple modules can be installed ata single site.
Conceptual design

Design status

1. Introduction
U-Battery is an advanced/small modular High Temperature Reactor (HTR), capable of providing a low-carbon,
cost-effective, locally embedded and reliable source of power and heat for energy-intensive industries andremote locations. The U-Battery concept is the result of a challenge initially set by Urenco to the Universityof Manchester and the Technical University of Delft to use existing, developed technology to bring nuclear toa market that, to date, has been serviced by diesel or other smaller sized fossil fuel or renewable technologies.
The universities were asked to consider all nuclear technologies that were available but to then discount thosethat would require long research programmes. The study confirmed that there were opportunities to design areactor that would be competitive when deployed at industrial sites and remote locations, and later developed
for further applications.
U-Battery is being developed by Urenco in collaboration with a range of delivery partners, including: Jacobs,
Cavendish Nuclear, and Kinectrics on reactor systems and safety, BWXT and NNL on fuel and fuel cycle,
Rolls-Royce and Howden on key components, Costain on construction and civil works, and Mammoet andDaher on transport. The project has achieved some significant milestones, notably winning a funding under
the UK Government’s Advanced Modular Reactor competition, announced in June 2020, and engagement with
the Canadian Regulator (CNSC) as part of their Vendor Design Review process. Development work to beundertaken in the frame of the AMR and VDR activities include design and development of the main nuclear
island components (reactor pressure vessel and cross-duct, intermediate heat exchanger, helium circulator,
control and instrumentation, and auxiliary systems for reactor cavity cooling, secondary shut-down, helium
purification) and conventional equipment including an aero-derivative turbine and generator set, process heatexchanger, and building/civil works.

2. Target Application
U-Battery is a multipurpose reactor consisting of a standardised reactor block (comprising the primary heliumand secondary nitrogen circuits, as well as a spent fuel store, and representing all of the nuclear-specific331systems and components), coupled to a user-specific interface that utilises the energy contained within thenitrogen coolant. Three broad user-specific applications are envisaged: a unit focussed on providing process
heat to a range of industrial applications; a unit focussed on electricity production via a gas-turbine; and a cogeneration unit capable of providing a variable mix of heat and power. Potential applications include: heat andpower supply to remote regions, localised heat and/or power supply for a range of existing industrial
applications, and potential new applications such as hydrogen production. In common with many nuclear
systems, rejected heat from the U-Battery primary applications could also be utilised for applications such as
district heating and desalination.

3. Design Philosophy
The U-Battery design aims to (i) employ only established and nuclear-validated technology, (ii) maximise theuse of modular manufacturing and off-site factory fabrication, (iii) be a cost-competitive and flexible energy
supplier, (iv) serve different regions and diverse markets utilising a common reactor island, and (v) support
alternative applications that may generate additional revenue or value. U-Battery exchanges the economies oflarge-scale for economies associated with localised power delivery (no transmission costs), simple safetysystems (made possible by the highly robust fuel, very small thermal output, high thermal capacity of themoderator, and efficient natural heat transport), and factory-based manufacture and modular construction(again made possible by its simplicity and small size).

4. Main Design Features
Reactor Core
The annular prismatic core is composed of 12 fuel columns, each composed of a stack of 5 hexagonal graphite
fuel elements, and arranged in a 3-lobed pattern. The fuel elements contain cooling channels for the helium
primary coolant, and closed channels containing cylindrical fuel compacts, each approximately 25 mm longand 12.5 mm in diameter. The central and radial moderating reflectors are formed from graphite blocks,
identical in geometry to the fuel blocks such that, if necessary, they can be replaced by the fuel handlingmachine during refuelling. Fixed graphite blocks provide a cylindrical core-former, and location is provided
by a core barrel and a restraint system similar to that used in UK gas-cooled reactors.

Reactivity Control
Fixed burnable absorbers made from B4C are used for through-cycle reactivity control and power shaping.
Control rods are provided to achieve shut-down under all plant states and to allow part-power operation.
Although not strictly necessary for nuclear safety, an independent shut-down system is also provided, basedon small absorber spheres that can be
gravity-fed into secondary shut-down
channels. These provide additional
regulatory confidence, and offer flexibility
during shut-down (for example, if needed,
control rods could be replaced early in thecycle without unloading the core).

Fuel Characteristics
The U-Battery fuel compacts contain
TRISO coated fuel particles, composed of auranium kernel coated in successive layersof pyrolytic carbon, silicon carbide, and anouter layer of pyrolytic carbon. The TRISO
particles are of a design recently validated
as part of the US-DOE’s AGR fueldemonstration program in the Advanced
Test Reactor at Idaho National Laboratory,
with uranium enriched to just under 20%.
Although the maximum fuel temperaturesin the U-Battery core never approach thewell-established safety limit for TRISO fuelwith UO2 kernels (1600oC), UCO kernels
validated in the Advanced Gas-cooled
Reactor (AGR) program will be employedbecause they offer even higher temperaturecapability, and were validated in greaternumbers.

332

Fuel Handling System
The fuel handling equipment for U-Battery is based on a well-proven pantograph system used in previous
HTRs (and fast reactors), together with a fuel carrousel arrangement based on experience with the UK’s gascooled reactors and on spent fuel handling experience at the UK’s Sellafield site. An on-site spent fuel store isprovided adjacent to the reactor core to facilitate core unloading prior to refuelling. The natural circulation
cooled spent fuel store is also based on experience with similar facilities at Sellafield. Much of the fuel handlingequipment is designed to be portable, and will not be left on site during periods of operation. This significantly
enhances the resilience of the facility against unauthorised access to fissile material, as well as allowing optimaluse of the refuelling equipment to service multiple U-Battery units.

Reactor Coolant System
In common with all previous HTRs, the U-Battery employs a helium primary coolant. The possibility of
employing CO2 was briefly considered early in the project, but irradiation experiments conducted as part of
the Manchester-Delft collaboration demonstrated that significant coolant chemistry intervention would benecessary to supress graphite oxidation, and this was not considered to be compatible with minimal (or evenno) operating personnel. The use of a secondary steam (Rankine) power cycle was rejected for similar reasons,and also because of the positive reactivity injection that would result from a heat exchanger failure and steam
ingress into the primary circuit. A direct-cycle configuration was rejected because of the significant
development requirements associated with a helium turbine. The preferred solution has been to adopt anindirect cycle configuration with a nitrogen secondary coolant, which can either pass heat to a tertiary fluid forheat applications, or drive a gas-turbine based on an aero-engine but without the combustion stage andemploying a closed-cycle configuration.

Reactor Pressure Vessel
The U-Battery core is housed in a Reactor Pressure Vessel (RPV) manufactured from SA-508/533 steel inorder to take advantage of the extensive experience from LWR RPVs. The RPV design is based on the ASMEIII design code for a working pressure of 40 bar at 300°C, together with limited transient temperature
excursions as permitted by the code. The helium coolant is passed to and returns from the primary heatexchanger via a coaxial duct housed within a cross-vessel, situated towards the bottom of the RPV reactor.
Limiting the design to a single major RPV penetration reduces the impact of a major failure of the cross-ductby eliminating the “chimney effect” that could result from multiple penetrations. The RPV wall is maintained
within acceptable temperature limits by the cool helium returned from the primary heat exchanger, which
passes up the annulus between the RPV and the core barrel before passing into an upper plenum and being
directed down through the core, into a lower plenum, and out to the heat exchanger through the inner part ofthe coaxial duct. The RPV is provided with a bolted head that can be removed for refuelling and inspection.

5. Plant Layout Arrangement
A plant layout has been developed based on a belowground reactor cavity and adjacent spent core storage
facility, both served by an overhead crane and collocated with the power conversion module in a vented
confinement building. The reactor and heat exchanger are positioned side-by-side, but each in its own
compartment. Both these are served by an above maintenance floor, that also allow refuelling and has accessto the used fuel cartridge store located next to the reactor cavity. Natural convection is used to cool the spent
fuel assisted with the special fuel store ventilation. The turbine generator is located in a separate building (onlyabove ground). The layout make provision for fuel to be loaded / unloaded into fuel casks with access via hefuel handling facility.

3336. Development Milestones
2008

2011
2017
2018

2019

2020
2023
2025
2028

334

The project was initiated by Urenco and the concept design was developed by the
Universities of Manchester and Dalton Institute in the UK and Technology University ofDelft in the Netherlands.
Feasibility study completed
Memorandum of Understanding signed with Bruce Powers in Canada
Green light to progress to Phase 1 of UK Government’s Advanced Modular Reactor
Programme
Completed the first stage of the evaluation process in Canadian Nuclear Laboratories’(CNL) invitation to site a first-of-a-kind small SMR in Chalk River, Ontario.
Established a service agreement with the Canadian Nuclear Safety Commission for prelicensing Phase 1 vendor design review.
U-Battery awarded funding by UK Government to conduct design and development tobring to market a new, innovative nuclear technology
Development of detailed design
Construction first-of-a-kind (FOAK) plant
First-of-a-kind U-Battery operatingAurora Powerhouse Product Line (Oklo
Inc., United States of America)
All content included in this section has been provided by, and is reproduced with permission of Oklo Inc, USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer, country oforiginReactor type

Oklo Inc., United States ofAmericaLiquid Metal Fast Reactor

Coolant/moderator

Liquid metal / no moderator

Thermal/electrical capacity,
MW(t)/MW(e)NSSS Operating Pressure(primary/secondary), MPa
Fuel type/assembly array4 – 150 / 1.5 – 50

Refuelling Cycle (months)

120 – 240 months

Design life (years)

40+ years

Plant footprint (m2)Less than 10,000 m2

Design status

At detailed design status.
Awarded a Site Use Permit by
the U.S. DOE as well as fuelmaterial by the Idaho National
Lab and the U.S. DOE.
Actively engaging the U.S.
NRC from 2016

Not pressurized
Metal fuel1. Introduction
Oklo is developing advanced fission powerhouses ranging up to 50 megawatts of electric power to provide
emission-free, reliable, and affordable energy. Oklo's powerhouses can produce clean energy for decades
without the need to refuel and have the capability to turn nuclear waste into clean energy.

2. Target Applications
The Oklo powerhouse is designed to provide affordable, reliable, emission-free electricity and heat. Oklo's
business model is to provide power on a power purchase agreement basis.

3. Main Design Features
Safety is fundamentally accomplished in Oklo's design by its inherent characteristics, including:- Low decay heat term, removed by inherent and passive means- Inherent reactivity feedbacks ensure reactor power is controlled during overpower or overtemperature
events- Multiple barriers to fission product release- Ambient pressure system removes sources of pressure and minimizes driving forces for release- Water not required for safety-related cooling

4. Initial Proposed Site
Oklo is working with a range of customers related to the initial sites. It is anticipated that the first powerhousewill be sited at the Idaho National Laboratory (INL) Site in southeast Idaho, referred to as the "Aurora INL
site." The Oklo Inc. site use permit request was evaluated by the Department of Energy Office of Nuclear
Energy (DOE-NE), a field office of the DOE, through the site use permit process and received a permit on
September 26, 2019. Additionally, the DOE provisionally approved a specific site within the INL Site for thelocation of the powerhouse in late 2021.

5. Plant Layout Arrangement
The Aurora site has a primary building with some supporting structures. The Aurora powerhouse has two
floors and has a footprint of an acre or less. The radiators are located external to the Aurora powerhouse. The
radiators exchange heat with ambient air as the ultimate heat sink for the power cycle. There is a parking lot
for the site as well as landscaping surrounding the site. Due to the inherent safety characteristics, an evacuation335

zone outside of the site boundary is not generally required.

Plant Layout

6. Design and Licensing Status
The design is at the detailed design stage. Oklo has completed fuel prototyping activities, heat transport testing
programs, and is now developing experimental advanced fuel recycling and fabrication capabilities. Oklo wasawarded a Site Use Permit by the U.S. DOE and was awarded fuel material by the Idaho National Lab and the
U.S. DOE. Oklo is the only company to have had a custom combined license accepted for review by the U.S.
NRC. Oklo has actively engaged the U.S. NRC from 2016 to the present in pre-application activities, as well
as application review activities.

7. Development Milestones
2016
2017
2018
2019
2019
2020
2021
2021
2021

2016-2022
2022
2022

336

Began pre-application interactions with the U.S. NRC
Demonstrated prototype fuel fabrication
Conducted thermal testing
Granted a site use permit from the U.S. Department of EnergyAwarded recovered fuel material from the Idaho National Laboratory
Combined license application submittal to the U.S. Nuclear Regulatory Commission
and acceptance for review (currently being revised)
Commercial customer partnership with Compass announced
Partnership with Centrus announced
Awarded U.S. Department of Energy Technology Commercialization (TCF) Fundaward related to fuel recycling
Awarded five awards from the Gateway for Accelerated Innovation in Nuclear for workon fuel data, fuel modeling, fuel fabrication, thermal hydraulic testing, licensing
support, and moreAwarded Advanced Research Projects Agency-Energy (ARPA-E) OPEN program
award related to fuel recycling
Awarded ARPA-E ONWARDS program award related to fuel recycling

HOLOS-QUAD (HolosGen LLC,
United States of America)
All content included in this section has been provided by, and is reproduced with permission of, HolosGen, USA.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
HolosGen, USA
country of origin
Reactor type
Coolant/moderator

High-Temperature Gas Reactor
Helium / Graphite

Thermal/electrical
capacity, MW(t)/MW(e)22 / 10

Primary circulation
NSSS operating pressure(primary/secondary), MPa

HeliumNo steam supply. Brayton Power Cycle
Max 7MPa, Min 3.5MPa
590 / 855Core inlet/outlet coolant
temperature (oC)
Fuel type/assembly array

TRISO-UCO hexagonal graphite elements

Fuel enrichment (%)
Refuelling cycle (months)
Core discharge burnup(GWd/ton)

19.95
96
61

Reactivity control
mechanismApproach to safetysystems
Design life (years)
Plant footprint (m2)
RPV height/diameter (m)RPV weight (metric ton)
Seismic design (SSE)

Redundant independent banks of Control
drums & Shutdown rods
Passive cooling40
30 (reactor comprised in ISO container)
5 (reactor fits 12.19 m long ISO container)
60 (Full Mobile Power Plant)
Designed for 3g max acceleration (all axis,
with active magnetic bearings)

Fuel cycle
requirements/approach

LEU, open cycle, spent fuel cartridges fit
transport and storage dry casks
Automatic, fitting shipping containers,
load-following, high efficiency.
Detailed design

Distinguishing features
Design status

1. Introduction
HOLOS-QUAD is a transportable high-performance gas-cooled reactor, with integral load following power
conversion systems housed within sealed power modules (SPMs), fully comprised within standard transport
containers. The design eliminates the need for the traditional BOP, greatly simplifies the design layout, reduces
capital, operational, maintenance and long-term spent fuel storage cost, while overall increasing safety andefficiency. A closed-loop full Brayton gas thermodynamic cycle converts the fuel cartridges thermal energy
into electricity at an efficiency of 45% with a load-following rate of 1MWe per minute. The whole microreactor
system is comprised within a single 40-foot (12.19-meter) reinforced container in compliance with ISO
dimensional requirements. The fuel is qualified, melt-resistant, TRISO fuel, graphite moderated, optimized toprovide 22 MWt (10MWe at 45% efficiency) in excess of 8 years before refueling, with an anticipated power
plant life span of 40 years.

2. Target Application
HOLOS-QUAD features high-temperature process heat that enables hydrogen production. The design features
load-following capability to match power demand, as a result, it can be employed as a distributable energy
source supplying power to microgrids or bulk power grids, and as a stand-alone electric island to, for example:i) charge electric vehicles; ii) supplement variable power compensating for the renewable energy sources; iii)
supply power to water-vessels' propulsions. The design scalability and utilization of integral power conversion
systems enables compactness for utilization as an emergency power generator rapidly deployable to disasterareas. Scaled-down versions of the design provide kW-class to MW-class generation for space applications.
337

Other applications include electric island capabilities to supply power to remote communities. Hardened
configurations have also been developed to satisfy military submerged, terrestrial and space applications.

3. Design Philosophy
Safety is paramount under all operating and storage modes: the design does not reach challenging temperatures
under design basis and beyond design basis accident scenarios, passively cooling the fuel under loss of coolant
scenarios and total loss of electricity. The core features a very low power density. Under worst case scenarios
the fuel maximum temperature briefly reaches approximately 1350°C, well below TRISO fuel melting
temperature, and does not release volatiles. The Brayton high-efficiency power conversion system ensures
minimum production of spent fuel per MW basis.

4. Main Design Features
Power Conversion
Closed loop Brayton cycle with intercooler
and recuperator heat exchangers integrated
altogether with the fuel cartridge, thus
eliminating the need for the traditional balanceof plant.

Reactor Core
Graphite moderated TRISO-fueled withburnable poison ensuring optimized burnout
with 8 Effective Full Power Years validated
for safe transport and operations.

Fuel Characteristics

Power Conversion System Test Assembly (HTA-1) for Brayton Cycle
Proof of Concept

Qualified TRISO fuel with UCO kernel.

Reactivity Control
The reactivity controls are represented by independent, diversified, and redundant reactivity control systems
based on sets of control drums and sets of shutdown rods.

Reactor Pressure Vessel and Internals
The design features 4 identical 5.5MW(t) SPMs, each equipped with Brayton power conversion components
in a configuration wherein the cycle pressure boundary is maintained by high pressure tubes. A simpler
configuration enables accelerated licensing by utilizing a nuclear-stamp RPV.

Reactor Coolant System
Core and auxiliaries are cooled by helium thermally coupled with the ultimate heat sink represented by air, or
water for terrestrial applications, air or water for water-vessel propulsion, and radiators for space applications.

Secondary System
The secondary system is only for process heat application, wherein helium circulates on one side of anintermediary heat exchanger and a user/application-selected working fluid circulates on the secondary side.

Steam Generator
The design does not need steam generators, however, steam can be generated as part of process heat for selected
applications.

5. Safety Features
Engineered Safety System Approach and Configuration
The design is equipped with traditional HTGR safety features. The design configuration represented by 4
identical SPMs features additional redundant pressure boundaries preventing migrations of radionuclides fromthe fuel cartridges (e.g., from potentially defective or damaged TRISO elements). In this design configuration,
the working fluid (helium) is only thermally coupled to the core: there is no physical contact between theworking fluid and core internals. The design relies on passive cooling under environmental extremes, loss of
coolant and loss of electricity so that the fuel temperature remains well below safety margins. The results ofanalyses on reactivity control aspects of the HOLOS-QUAD design under load-following operations published.

Decay Heat Removal System / Reactor Cooling Philosophy
The core is represented by a relatively large mass of graphite with respect to the total amount of fuel, as a result
the fuel does not experience temperature spikes under total loss of coolant circulation and total loss of coolant
scenarios. The validation on thermal-hydraulic and neutronics performance, as well as the effectiveness of
passive decay heat removal of the HOLOS-QUAD have been published by the Argonne National Laboratory.
338

Spent Fuel Cooling Safety Approach / System
Spent fuel cartridges are inserted inside relatively small dry casks. Cooling is executed passively by naturalconvection with the air surrounding the shielded dry cask.

Containment System
The design utilizes TRISO fuel which features "functional containment": radionuclides resulting fromoperations are trapped within TRISO microspheres. In addition to the functional containment, the design
features two additional pressure boundaries that prevent radionuclides migrating from potentially defective or
damaged TRISO microspheres to the working fluid, into the environment.

Chemical Control
The design eliminates the balance of plant normally dedicated to filter and makeup the working fluid, and theredundant active magnetic bearings eliminate "metal-to-metal" contact between rotary and stationary
components, even during transport.

6. Plant Safety and Operational Performances
The design follows the characteristics of HTGRs and features a very low power density core and a very highheat capacity. These characteristics result in slow fuel temperature variations, which under worst case scenarios
do not challenge the high temperature fuel loaded in the reactor under all postulated off-normal and accident
scenarios.

7. Instrumentation and Control System
The design follows the electrical load automatically and features a control system that regulates the power
flows between the thermal-to-electric conversion components, the ballast/black-start batteries, and theelectrical load. Process instrumentation and control systems regulating the active magnetic bearings, control
drum and shutdown rod drives, and the compressor speed, enable fast-acting power maneuvering during
normal operations. The digital instrumentation and control equipment is located within a 20-foot “digital I&C”
ISO container electrically coupled to the 40-ft ISO container via shielded, EMI- and EMP-shunted data andpower cables.

8. Plant Layout Arrangement
The design features a highly integrated,transportable nuclear power plant fullycomprised within the dimensional constraints ofa standard shipping container with feasibility
validated by the Argonne National Laboratory.
For above surface applications, the ISO container
is positioned via tractor trailer inside a
prefabricated non-permanent concrete “shieldbuilding”. For underground operations, the ISO
container can be “verticalized” (the design's
active magnetic bearings enable positioning of
the container horizontally, angled, or vertically).
The vertical ISO container can be positioned
within a shielding bored hole with aprefabricated concrete liner and passive aircooling pathways.

Pre-fabricated non-permanent shield building and dose rate
calculations at the 50-m fence

9. Testing Conducted for Design Verification and Validation
The design has been scrutinized for approximately 4 years by design and resource teams formed by scientists
and subject matter experts at the national laboratories, industry and academic institutions, sponsored in part by
the U.S. DOE ARPA-E MEITNER funding program. Validation activities under this program included
technical and economic feasibility executed with high-fidelity codes developed by the national laboratories.

10. Design and Licensing Status
Preliminary safety performance analyses have been completed and verified through operations of the SPM
subscale simulator (SPM-SS) configured to mimic design basis accident scenarios. Preliminary safetyassessments through analyses executed with the cooperation of teams at the national laboratories and results
validated via the electrically-heated SPM-SS were completed in 2022. A formal safety assessment request isplanned for submission to the U.S. Nuclear Regulatory Commission in 2023.

11. Fuel Cycle Approach
The 10MWe HOLOS-QUAD design is equipped with a core optimized for 8 EFPY. Spent fuel cartridges
passively dissipate decay heat from reactor shutdown, and can be placed within dual purpose transport andstorage shielding dry casks within 72 hours from shutdown. The spent fuel cartridge dimensions altogether
339

with those of the dry cask are designed to comply with ISO shipping container dimensional constraints. The
spent fuel cartridges will be stored at long-term storage facilities and the spent fuel in them will not beprocessed as the type of fuel utilized is already in a stable form for long-term storage.

12. Waste Management and Disposal PlanAs the coolant in this design is not physically in contact with the materials forming the fuel cartridge internals,
there is no working fluid contamination. As a result, the great majority of the power conversion components
and their ancillary components are not exposed to irradiation effects (e.g., particulate circulating through thepower conversion components) and will not require specialized decommissioning procedures. These
components will be classified as non-radioactive waste, cleaned and refurbished, or disposed of. Cost of spent
fuel handling, transport and storage for 50 years is included in the HOLOS-QUAD costing analyses.

Sealed Power Module Subscale Simulator (SPM-SS) Validation Test Assembly

SPM-SS with helium at 7MPa

13. Development Milestones
2015
2018
2021

2022

340

Conceptual design for F&A Holos Reactor.
Demonstrated feasibility of the power conversion system using the HTA-1 test assemblyAwarded the DOE ARPA-E MEITNER funding to validate HOLOS-QUAD design throughmodelling, simulation and testing.
Awarded the DOE-NE GAIN Voucher with Argonne National Laboratory to develop innovative
material solutions to enhance core performance.
Completed construction of the Sealed Power Module subscale simulator (SPM-SS) with helium
as working fluid at 7 MPa pressure for the full-scale SPM.
Completed design validation demonstrating: 45% net plant efficiency; fully fitted within a single
40-foot ISO shipping container; 8 EFPY, passive decay heat removal; and load-following
capability at a rate of 1MWe/min. Key design aspects validated via testing.

MARVEL Research Microreactor (Idaho
National Laboratory, USA)All content in this section has been provided by and is reproduced with permission of Idaho National Laboratory, USA.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer, country
Idaho National Laboratory, USAof origin
Reactor TypeLiquid Metal Cooled Thermal reactor
Coolant/Moderator
Sodium-Potassium Eutectic, Hydrogen in fuelThermal/electrical capacity,
0.075-0.1 / 0.015-0.027
MW(t)/MW(e)Primary Circulation
Natural Convection
NSSS Operating Pressure
0.39 / 0.22
(primary/secondary)Core Inlet/Outlet Coolant
480 / 548
temperature (C)Fuel Type/Assembly array
Uranium Zirconium Hydride
Number of Fuel Assemblies in6
the core
Fuel Enrichment
19.75%
Refueling cycle (months)> 60Core Discharge Burnup
2.098 (current; can be increased if needed)(GWd/ton)
Reactivity Control Mechanism
Four ex-core, safety-related control drums, One
defense-in-depth central shutdown rod
Approach to safety systems
Doppler-based inherent safety in fuel; Primarycoolant, intermediate coolant, and decay heatremoval is conducted via passive, naturalconvection. Loss of on-site power actuates
shutdown mechanism passively
Design Life (years)
2-40 (depending on transients)
Plant footprint (m2)8.9
RPV height/diameter (m)2.55 / 1.22
RPV weight (metric ton)
5.94
Seismic Design (SSE)
IBC 2015 Risk Category IV Site Class C
Fuel Cycle
Has refueling capability, but its current system
Requirements/approachlife is limited to 2 years
Distinguishing Features
Compact accessible core; naturally driven,
integrated power conversion systems; adjustablehigh-grade heat extraction; low-grade heat
extraction capability; small footprint
Design Status
Equipment manufacturing in progress

1. Introduction
The Microreactor Applications Research Validation and Evaluation (MARVEL) Project entails the design,development, construction, and startup of an INL test microreactor, funded by the United States Departmentof Energy (DOE) via the Microreactor Program (MRP). The goal of the project is to establish an operational
nuclear applications test bed that can generate combined heat and power to enable integration and R&D
with end-user technologies, as well as allow microreactor technologists to test next-generation control
systems. The microreactor is a thermal reactor utilizing Uranium Zirconium Hydride (UZrH) fuel withreview and authorization by the Department of Energy Idaho Operations Office (DOE-ID) for National
Environmental Policy Act (NEPA) compliance, safety review, and supplemental readiness assessments forstartup and operation. To enable rapid deployment, the MARVEL reactor will reside in the Transient
Reactor Test (TREAT) Facility and utilize the existing operating Category B reactor facility, approved
facility safety basis, operating crews, and recent re-start experience. The MARVEL team consists of Idaho

341

National Laboratory (INL), Argonne National Laboratory, Los Alamos National Laboratory, WalshEngineering, Qnergy, Munro & Associates, and Creative Engineers Inc.
The primary purpose of investigating a near-term concept is to determine its feasibility and utility for future
microreactor research and development, such as a test bed for integrating microreactors with electric andnon-electric applications and demonstrating the technology to key end-users. It can also drive infrastructuredevelopment and capabilities to support future microreactor demonstrations.
The U.S. DOE Microreactor Program supports research and development (R&D) of technologies related tothe development, demonstration, and deployment of very small, factory-fabricated, transportable reactors
to provide power and heat for decentralized generation in civilian, industrial, and defense energy sectors.
Such applications currently face economic and energy security challenges that can be addressed by this newclass of innovative nuclear reactors. Led by INL, the program conducts fundamental and applied R&D to
reduce the risks associated with new technology performance, manufacturing readiness, and deployment of
microreactors. The program aims to verify that microreactor concepts can be licensed and deployed bycommercial entities to meet specific use case requirements. Microreactors, often referred to as specialpurpose reactors or very small modular reactors, are factory manufacturable, easily transportable, anddesigned to produce tens of megawatts of thermal (MWth) energy. This power limit allows microreactors
to be classified as Hazard Category 2 per the Code of Federal Regulations (CFR) at 10 CFR 830 and DOESTD-1027. These reactors are decentralized energy sources that can provide sustainable and affordable heatand power to remote communities and industrial users while having self-contained geometry that requiresvery low maintenance. Microreactors are intended to be self-regulating and do not rely on engineered
systems to ensure safe shut down and removal of decay heat.

2. Target Application
MARVEL is targeted at researchers, technology developers, and end-users for enabling the successful
deployment and adoption of microreactors through research and education. Some examples of R&D
activities are:Test, demonstrate, and address issues related to installation, startup, and operation
 Simplify siting and environmental review process Startup methodology for microreactors
 Normal operating transients such as startup and load management
 Cyber and physical security hardening Seamless integration to a net-zero, electrical microgrid Demonstration of high and low-grade heat extraction.
Enable Autonomous Operation Technologies
 Automate operator functions, while maintaining reactor safety
 Demonstrate radiation and temperature-hardened sensors and instrumentation to enable remotemonitoring, advanced sensor reliability tests, and online calibration
 Live data can feed a digital twin of the reactor to “train” an artificial intelligence-based control
system Demonstrate wireless transmission of live data of both electrical and thermal power output during
startup, operation, and shut down. This allows real-time feedback on system output, performance,and prediction of any unplanned maintenance needed in an operating microreactor.
Enable Seamless Application Integration
 The control systems manage the energy grid demand and reactor power and heat supply. This
management requires a carefully designed control system that can predict the interplay ofcontrols, thermal inertia, and reactivity feedback.
 Demonstrate integration approaches for a range of applications investigating both reactor power
management and load management approaches342

3. Design Philosophy
The design philosophy is to select technologies that can lead to criticality within the shortest time possible
and not necessarily on the technology readiness level. Time is the essential metric of project performance
without sacrificing safety and quality. This drives the team to select technologies that are either (i) off-theshelf, (ii) can be implemented, and/or (iii) can be developed to accelerate technology and manufacturing
readiness levels. Innovation and creativity are embedded in the team culture and celebrated. This includes
adopting multiple engineering tools and processes from non-nuclear technology industries, e.g. Scrum
Agile Product development, and Design for Manufacturability (DFM) modeling from the automotive
industry.

4. Main Design Features
Power Conversion
MARVEL utilizes commercial, off-the-shelf Stirling Engines from Qnergy. Each Stirling engine canproduce 5-7 kilowatt and comes equipped with supporting ancillary equipment for low-grade heat rejection.
Qnergy equipment will absorb available high-grade heat and then convert that heat into either using electric
power or low-grade heat for rejection into site processes or finally into the ultimate ambient heat sink. An
alternate high grade heat extraction system will be available for extracting high grade heat for process heatapplications testing.

Reactor Core
The MARVEL reactor will contain fuel elements sourced from TRIGA International (TI): LEU, 30%
uranium, by weight, 19.75 % enriched (TRIGA LEU 30/20). Beryllium and graphite reflected, ZrH
moderated, solid fuel, tank type reactor. The MARVEL reactor core comprises 36 LEU fuel elements
arranged in a triangular pitch lattice around a central voided location for the central insurance absorber(CIA) rod assembly. Surrounding the core is a thick axial neutron reflector composed of beryllium oxide.

Fuel Characteristics
The major components of the MARVEL fuel system are (1) U-ZrH1.6 fuel slugs (annular), (2) Zirconium
filler rods (to be inserted into fuel slug annulus), (3) Molybdenum spacer disk placed between lower
graphite reflector and bottom of the fuel stack, (4) Stainless steel 304 cladding, (5) Top and bottom graphite
axial reflectors, and (6) Top and Bottom end plugs.

Reactivity Control
The MARVEL reactor also has a very high net negative temperature feedback for prompt reactivity control
inherently due to the Doppler broadening of resonances. Manual and Passive Reactivity Control andshutdown are achieved with four custom-designed, safety-related Control Drums (CDs) and a defense-indepth Central Insurance Absorber (CIA) that control reactivity via motor drivers in the control cabinet toposition the drums and rods within the reactor structure. Passive actuation functions are built into the designfor loss of power and inadvertent energizations of the motors.

Reactor Pressure Vessel and Internals
The Primary Coolant System (PCS) is a high-temperature, low-pressure boundary that houses the coreinternals, reactor primary coolant, and argon gas headspace. In addition, the PCS passively maintains decay
heat removal capability. The boundary is a metal weldment made from 316H stainless steel for hightemperature reactors designed per ASME Section III Division 5.

Reactor Coolant System
The primary coolant system (PCS) is a four-loop hydraulic circuit assembled to transport nuclear fission
heat from the nuclear fuel to the intermediate heat exchanger (IHX) using the natural convection flow of
the primary coolant. The PCS also transfers decay heat to the ultimate heat sink. Approximately 120 kg of
NaK, liquid metal at room temperature, serves as the primary coolant. The Stirling engine heat exchangers
connect to the reactor vessel and interface with the NaK coolant via the IHX containing lead. The Stirlingengine coils or high-grade heat exchanger, depending on configuration, extract heat from the primarycoolant and reduce the NaK temperature. The cooled and denser NaK then flows outward to the periphery343

of the core, then downward through four downcomer pipes outside the beryllium oxide side reflector andthrough in the lower plenum. The NaK then rises back up through the active core under natural circulation
forces driven by the heated section of the active core.

Secondary System
Lead has been selected as the coolant for the IHX, serving
as the secondary system. An argon gas sweep will bemaintained on the IHX overhead gas space to remove any
volatile activation products for ease of maintenance andrepair.

5. Safety Features
MARVEL employs a high safety pedigree for (i) reactivity
control, (ii) safety shutdown, and (iii) decay heat removalby relying primarily on physics rather than engineered
systems. The fuel has a high prompt net negative
temperature reactivity feedback due to the strong Doppler
broadening of resonances by having hydrogen close to thefuel. In addition, passive shutdown features enable the use
MARVEL Integral Effects Test systems,of potential energy to actuate negative reactivity in the core.
structures, and components
Both primary and intermediate coolants are driven by
natural circulation. Decay heat is removed from the fuel to the outer shell of the reactor primarily byconduction and radiative heat transfer and subsequently rejected to ambient air by convection. Finally, there
are three robust fission product barriers- fuel cladding, primary coolant boundary, and top & bottom guard
vessel to confine hazardous materials throughout system life in all operational, postulated design-basis, andbeyond-design-basis events.

6. Testing Conducted for Design Verification and Validation
The MARVEL team conducts rapid prototyping tests to mature its technologies. So far, more than ten
separate effects tests have been conducted, including, but not limited to, the intermediate heat exchangers,
control drums, instrumentation control, neutron detection, shutdown rod actuators, and Stirling engines.
Due to MARVEL’s novel thermal hydraulic circuit, where a liquid metal natural circulation primary loop
is in series with four parallel liquid metal natural circulation loops, an integral effects test (IET) of thesystem was considered necessary for verifying the transient dynamics of the system before reactor
construction. Hence the team has successfully designed and fabricated a full, scale electrically heatedprototype of the MARVEL reactor. This is conducted per ASME boiler and pressure vessel code (Section
III and VIII) within nine months by employing an agile development process. The test hardware includes(i) a Full-scale mechanical IET test article; (ii) eight electrical control cabinets; (iii) a structural frame; (iv)four IET flow meters; (v) more than 200 thermocouples and pressure transducers; and (vi) four Stirlingengines, engine control, and heat rejection units.
The goals of this IET are to: (i) validate flow and heat transfer characteristics of MARVEL technology; (ii)benchmark modeling & simulation parameters; (iii) streamline manufacturing methods; (iv) de-risk supply
chain; and (v) train operators.

7. Design and Licensing Status
DOE conducted an environmental assessment (EA) as part of the NEPA process, which analyzed thepotential environmental impacts of constructing the MARVEL microreactor inside Idaho National
Laboratory’s (INL’s) Transient Reactor Test Facility. At the conclusion of the EA process, DOE issued afinal EA with a “finding of no significant impact” (FONSI) for the MARVEL Project (DOE/EA-2146). A
safety design strategy (SDS-119, Rev 1) has also been approved by DOE. The MARVEL reactor is currently
in Final Design Phase, with 90% construction planned to conclude in Q2, CY23, and fuel load and initial
criticality in Q2, CY24.

344

MMR® (Ultra Safe Nuclear Corporation,
United States of America)
All content included in this section has been provided by, and is reproduced with permission of Ultra Safe Nuclear Corporation, USA.

Parameter

MAJOR TECHNICAL PARAMETERS
Value

Technology developer,
country of origin
Reactor type

MMR Unit

Coolant/moderator

High Temperature Gas-cooled Reactor /micro-reactor / nuclear battery
Helium / Graphite

Thermal/electrical
capacity, MW(t)/MW(e)Primary circulation

15 / >5 and30 / >10
Forced circulation

NHSS Operating Pressure(primary/secondary), MPa
Core Inlet/Outlet Coolant
Temperature (oC)
Fuel type/assembly array3 / 0.1

Fuel enrichment (%)

Silicon Carbide

Core Discharge Burnup(GWd/ton)
Refuelling Cycle (months)
Reactivity control
mechanismApproach to safetysystems

TRISO
Particles

Fully Ceramic Micro-encapsulated (FCM™) fuelUltra Safe Nuclear Corporation, USA

Helium 300 / 630, solar salt stored at
560
FCM® (TRISO based) graphite /Hexagonal
HALEU 19.75%> 60
240 (20 years)

Design life (years)

Control rod insertion, negative
temperature coefficient.
Passive (Category A), no moving parts,
does not require fluid or naturalconvection
20 (fuel core), 40 (vessel)

Plant footprint (m2)

80 x 80 (citadel area)130 x 96

RPV height/diameter (m)RPV weight (ton)

13.25 m (including lid and stand pipes)/ 3.5 m30.1 metric tonnes

Seismic Design (SSE)

0.3g

Fuel cycle requirements /ApproachDistinguishing features

20 year core life, can be refuelled for an
additional 20 year cycle
No core meltdown; modular reactor andmodular powerplant, adjacent nonnuclear power conversion plant; no
EPZ required; load following / fully
dispatchable; nuclear reactor isolated
from load via molten salt loop; < 6months assembly at site
Basic Design, project in formal
licensing with the regulator in Canada

Design status

1. Introduction
The MMR® system is a small modular nuclear energy system that delivers safe, clean and cost-effective
electricity and heat to remote mines, industry and communities. The energy system consists of two plants, thenuclear plant and the adjacent non-nuclear power plant. The Nuclear Plant is independent of the Adjacent Plant,
requiring no supporting services for any event for its safe operation. The Adjacent Plant consists of theequipment and systems that convert the process heat to electrical power or other forms of energy as per client
requirements. The Nuclear Plant generates approximately 15 MW(t) of process heat that can supply electrical
power and/or heat the end users. The electrical power can also be supplied to an area grid. The fuel life is 20
years with the possibility of refuelling for an additional 20-year cycle.

345

2. Target Application
The MMR power plant is designed for remote communities as standalone micro grid or for heavy industry
applications such as process heat and hydrogen production. MMR energy system can scale to match demandwith multiple units at the same site.

3. Design Philosophy
The design philosophy of the MMR uses the following main principles: (i) Use proven and existing technology;(ii) Design for simplicity; (iii) Design for reliability; (iv) Separate the standardized nuclear plant from theapplication adjacent plant; (v) Design for low maintenance; (vi) Design for modular construction andreplacement, and (vii) Design all components to be replaced in principle. The design builds on the High
Temperature Gas Cooled Reactor knowledge base and is constrained by known material limits as determinedby design codes and standards. The reactor is fuelled to operate for 20 years with the possibility of refuellingfor an additional 20-year cycle.

4. Main Design Features(a) Power Conversion
Electricity is generated in the power generation system located in the adjacent plant from the heat supplied
from the nuclear plant via the molten salt. The power generation system consists of the turbine generator andsupporting infrastructure. The adjacent plant has a main electrical grid connection for supply of the electrical
power generated via transmission infrastructure.

(b) Reactor Core
The reactor core consists of hexagonal graphite blocks that contain stacks of the Fully Micro Encapsulated(FCM) fuel pellets. The MMR reactor core has a low power density (1.24 W/cm3) and a high heat capacity.
This results in very slow and predictable temperature changes. The MMR reactor is fuelled once for its initial
20-year operating life and sealed. The core can then be replaced to extend the operating life of the reactor for
another 20-year cycle. The core provides adequate coolant flow paths for heat removal, and the graphite
material of the blocks assists with further heat removal.

(c) Fuel Characteristics
The MMR reactor uses Ultra Safe Nuclear Company proprietary Fully Ceramic Micro Encapsulated (FCM)
fuel. FCM fuel provides a new approach to inherent reactor safety by providing an ultimately safe fuel. FCM
is based on very small particles containing low-enriched uranium. FCM fuel ensures containment ofradioactivity during operations and accident conditions, which means that no fission products are released fromthe fuel.
FCM is manufactured with industry-standard Triple Coated Isotropic (TRISO) fuel particles, whose primary
purpose is to retain fission products. The TRISO fuel, which contains the radioactive by-products of fission
within layered ceramic coatings, is further encased within a fully dense silicon carbide matrix. This is likeencasing the fuel in a diamond-like substance. The combination of TRISO fuel particles and the silicon carbide
coating provides an extremely rugged and stable fuel with extraordinary high temperature stability.

(d) Fuel Handling System
The MMR is initially fuelled for a 20-year cycle, and may be refuelled for an additional 20-year cycle. Thereis no long-term fuel storage on site.

(e) Reactivity Control
The core provides for areas for insertion of control rods. The MMR reactor core has a low power density anda high heat capacity resulting in very slow and predictable temperature transients. The core has strong negative
temperature feedback.

(f) Reactor Pressure Vessel and Internals
The Reactor Pressure Vessel (RPV) houses the core. The MMR has a drain at the base of the Reactor Pressure
Vessel and avoids major water ingress events because there is no steam generator in the primary circuit. The
MMR Reactor Pressure Vessel is ‘helium tight’. The MMR design includes metallic seals and a seal weld toensure leak tightness. The Reactor Pressure Vessel is designed in accordance with the ASME Boiler & Pressure
Vessel Code (BPVC).
The colder helium flows up against the reactor vessel wall to avoid overheating from the core outlet gas. The
MMR employs a design principle in the primary circuit that ensures that the hot helium is always surroundedby cold helium at a higher pressure. This means that any leaks are from cold to hot and will eliminate thepossibility of hot gas impinging on the pressure boundary.

346

The MMR design studies the effect of core bypass flow by modelling extreme and bounding cases. The MMR
design avoids problems with flow across tube bundles in the intermediate heat exchanger by using printed
circuit technology that does not have any tubes.

5. Safety Features
The MMR technology has been developed by Ultra Safe Nuclear Company and is based largely on provendesigns with inherent safety features. The safety design is then further augmented with specific novel safety
features. The use of proven inherent safety design features confers confidence in the operability and safety of
the facility, while the novel safety features enhance the safety of the technology. Operations are simple withminimal operations and maintenance requirements, and no on-site fuel storage, handling, or processing.
The MMR reactor is a truly walk away, safe power reactor. In the case of a fuel or cooling system failure, theMMR reactor cannot melt down as the temperature will remain at a safe level while heat dissipates passively
into the environment.

(a) Engineered Safety System Approach and Configuration
The MMR safety functions are achieved through a combination of inherent safety features and engineered
safety features. The inherent safety features result from the selection of materials used in and design featuresof the reactor fuel and core, moderator and coolant. Engineered safety features are those introduced specificallyto perform a safety function, and they may include both passive and active structures, systems and components(SSC).

(b) Decay Heat Removal
The reactor will maintain a safe temperature indefinitely even in the event of a cooling failure because of thefuel characteristics and the reactor design. In a cooling failure, the reactor is cooled through passive heattransfer. External water, electricity, and additional active systems to dissipate heat during a system failure are
not required.

(c) Chemical Control
The inherent features for controlling chemical attack of the fuel of the MMR design include:•

Design features that limit water ingress•

The circulator magnetic bearings eliminate the source of water that was a major contributor to the lowcapacity factor of the Fort St. Vrain HTGR. The primary heat transfer cycle and steam generation
cycle has been separated with the addition of a low-pressure intermediate loop.

•

The primary cycle which is pressurized to limit air ingress•

The reactor core graphite will not ignite if were to come into contact with air even at operating
temperatures• The primary cycle coolant (helium) is chemically inert•

The use of ceramically-coated TRISO as a basis of the fuel•

The full ceramic microencapsulation of the TRISO fuel particles to further control chemical attacks•
The use of a nuclear-grade vessel(d) Containment
The MMR’s fuel performs the function of containing fission products during such conditions. This sets the
MMR apart from most current operating reactor technologies, which rely on highly specialized and complex
safety systems to prevent and mitigate further releases of fission products that escape their fuel in the case of
postulated accidents. Retention of fission product is achieved by the FCM fuel particles. The fuel safety margin
is so large that fission product retention is accomplished entirely by the fuel. No other containment is needed.
The reactor core is sealed for life – the fuel cannot be accessed.

6. Plant Safety and Operational Performances
The MMR technology has been developed by Ultra Safe Nuclear Company and is based largely on provendesigns with inherent safety features, further augmented with specific novel safety features. The degree of suchproven inherent safety design features confers confidence in the operability and safety of the facility, while thenovel safety features further enhance the confidence in the technology. Operations are simple with minimal
operations and maintenance requirements.

347

7. Plant Layout Arrangement
The MMR facility includes a Nuclear Plant containing
an MMR reactor, and an Adjacent Plant, which are themain physical works related to the Project.

Nuclear Plant
The Nuclear Plant provides process heat to the
Adjacent Plant where it is converted to electrical power
and/or heat as per client requirements. The Nuclear
Plant contains the Nuclear Building and the Citadel
Building, where the MMR reactor and its associatedNuclear Heat Supply System are housed. This building
protects the reactor and the Intermediate Heat
Exchanger from hazards (both external and internal to
the Citadel Building), and the wall provides biologicalshielding.

Adjacent Plant

MMR two-unit Power Plant

The Adjacent Plant buildings contain the equipment required for the generation of electricity from the heat
supplied by the Nuclear Plant and to interface with any customer end-use facilities. The Adjacent Plant Molten
Salt Heat Storage System acts as an intermediary to transport and store the heat generated in the Nuclear Plant
and transfer it to a steam cycle for the purpose of generating power and heat for customer applications.

8. Design and Licensing Status
Global First Power (GFP) has submitted a ‘License to prepare site initial application’ for the MMR Demopower plant at the Chalk River site to the Canadian Nuclear Safety Commission (CNSC) which regulates all
nuclear activities in Canada. Formal licensing process is underway for the MMR Project at Chalk River, in
Canada. The MMR commercial demonstration project has a Project Hosting Agreement with an identified site
at Chalk River Laboratories in Canada. The formal licensing process is in the first phase (License to PrepareSite), which will be followed by the License to Construct phase.

9. Fuel Cycle Approach
The MMR will be fuelled to operate for 20 years. Refuelling is possible to enable a second 20-year cycle, withthe balance of plant designed for a 40-year life. The Decommissioning phase activities are anticipated to takeapproximately 6 months. Spent fuel will be stored according to the national storage plan. No processing or
conditioning of the spent fuel is required because FCM fuel is already in a geologically stable form.

10. Waste Management and Disposal Plan
All waste will be handled and processed in a responsible and safe manner that ensures minimum exposure to
all personnel handling, transporting and processing the waste. Waste will be segregated at source as nonradioactive waste and radioactive or potentially radioactive waste. Waste will be temporarily stored on the
Project site in defined areas and transported to authorized processing facilities at appropriate times, dependent
on the category and type of waste.

11. Development Milestones
Ultra Safe Nuclear Company MMR Demo unit at Chalk River (Canadian Nuclear Laboratories)
2011

Secured FCMTM fuel and MMRTM reactor patents

2016

Established R&D and fabrication laboratories

2017

Initiated FCMTM fuel development and qualification

June 2018

GFP submits proposal to CNL, supported by OPG & Ultra Safe Nuclear Company

February 2019

CNL announces GFP proposal enters Stage 3 proponent review process

March 2019

GFP submits license to prepare site initial application to CNSC

2020-2023

License to prepare site, complete environmental assessment, and begin site preparation

2021-2027

Site Preparation and Construction

2026-2052

Plant Operation

2052-2060

Decommissioning and abandonment

348

Westinghouse eVinci™ Micro Reactor(Westinghouse Electric Company LLC, USA)All content included in this section has been provided by, and is reproduced with permission of, Westinghouse Electric Company LLC, USA.

MAJOR TECHNICAL PARAMETERS
Parameter
Value
Technology developer,
country of origin
Reactor typeWestinghouse ElectricCompany LLC, USA
Heat pipe cooled

Coolant/moderator

Heat pipes / metal
hydride moderator
7 – 12 / 2 – 3.5

Thermal/electrical capacity,
MW(t)/MW(e)Primary circulation
NSSS Operating Pressure(primary/secondary), MPa
Core Temperature (oC)
Fuel type/assembly array

Heat pipes
N/A
NA800

Number of fuel assemblies

TRISO or anotherencapsulation
Monolith core

Fuel enrichment (%)

5 – 19.75

Core Discharge Burnup(GWd/ton)
Fuel cycle (months)

Not Disclosed

Reactivity control
mechanismApproach to safety systems

Ex-core control drums> 36

Design Life (years)

Inherent and passive
safety for shutdown andheat removal
40

Plant footprint (m2)

< 4 000

RPV height/diameter (m)N/ASeismic design (SSE)

IBC Zone 4 Category F

Fuel cycle requirements /Approach

No onsite refuelling
Replace reactor
approachTransportable reactor
that can operates
autonomously
Conceptual design

Distinguishing features

Design status

1. Introduction
The Westinghouse eVinci Micro Reactor is designed for energy generation in remote or isolated locations. The
design can produce both process heat and electricity for remote communities, mining operations, or criticalinfrastructure installations. The key attribute of the design is its transportability within standard shipping
containers. The design is based on heat-pipe reactor technology that has been developed and tested by Los
Alamos National Laboratory for space applications. Because of its compact and simplified design, the eVinci
Micro Reactor will be manufactured and fuelled in a factory, and then transported to an end user site. The
figure below shows how the eVinci reactor and power conversion system can be packaged into two standardtransport containers. One of the containers houses the reactor and the power conversion system. The othercontainer includes power electronics and the instrumentation & controls (I&C) system.

349

Reactor Canister

Instrument and Controls
Power Electronics
Brayton Power Conversion
Westinghouse’s eVinci Micro Reactor schematic.

2. Target Application
The eVinci Micro Reactor is designed specifically to serve remote communities, mining operations, or military
installations. It combines both heat and power generation capabilities, addressing diverse energy needs of thesedecentralized and off-grid markets.

3. Design Philosophy
The design of the eVinci reactor leverages proven heat pipe technology developed by the Los Alamos National
Laboratory (LANL) for space application. This uranium-fuelled reactor does not use a bulk primary coolant.
Instead, heat is removed from its core using passive heat pipes, limiting the number of its moving parts andproviding overall plant simplicity. The design utilizes the inherent safety features in the fuel, moderator andheat pipes to enhance safety and self-regulation capability.

4. Main Design Features
The fundamental reactor design is based on a solid core block that includes a matrix of nuclear fuel, moderator
and heat pipes that extract the heat from the core region. The reactor core by itself is subcritical. It requires
radial and axial reflectors to improve neutron utilization. Reactivity control and shutdown are performed withradial control drums placed around the core. Shutdown can also be performed with shutdown rods that can beinserted into the core block. The control drums are the only moving parts within the reactor canister leveraging
the passive heat pipe thermal exchange. The reactor heat transport design utilizes the unique properties of
sodium heat pipes. Sodium heat pipes are embedded in the solid core and under normal operation are used tomove the core heat to the heat exchanger and associated power conversion system. Heat pipe operation
removes the need for coolant pumps in the primary system. The complete reactor and power conversion
systems are controlled with an autonomous control system that is based on proven nuclear I&C systems.
The eVinci Micro Reactor has inherent safety features. For example, in case of decay heat removal, the core
design within the reactor canister enables heat removal by conduction to the outer sections of the reactor
canister that passively allow heat transfer to the surrounding atmosphere (air). The core is designed withnegative reactivity feedback, allowing for increased safety in the event of an accident scenario. If external
power is lost or the reactor is tripped through its autonomous control system, the control drums automaticallyrotate to a high neutron absorption position around the core and subsequently shutdown the reactor. This solidcore design eliminates many traditional accident scenarios such as high-pressure pipe ruptures or loss of
coolant accidents.
An open-air Brayton system was incorporated due to its compactness, technical maturity and reliability. A
single shaft gas turbomachinery composed of a compressor, turbine, and alternator operating as a single rotary
spool. The high-speed single-shaft architecture results in a very compact power module. An annular recuperateand simplified, low pressure ducting will enable close integration of the engine with the primary heatexchanger. To avoid the use of lubricants, magnetic bearings will allow the engine to operate in any orientation,
facilitating optimal integration and packaging. The high-speed alternator acts as a motor to start the engine
during regular and black start up. The mechanical configuration is critical to starting the engine with therelatively slow reactor ramp-up and avoidance of thermal stress. Power electronics are combined with thevariable speed engine to enable efficient part-load and load following power management.

350

5. Resilient Features
Resiliency of eVinci Micro Reactor is enabled through packaging within a secure canister, installed in a strong
secure vault on site and connected to a micro-grid coordinating both heat and power. A dedicated device called
SMART is incorporated within its power conversion and heat delivery systems, along with its autonomous
load following capability make it ideal to operate together with other energy resources. The integration
capability of the eVinci Micro Reactor to connect to a micro-grid can make users resilient by utilizing thediversity, high-reliability and safety of nuclear energy with other power generation resources.

6. Operational Performances
The eVinci Micro Reactor leverages its autonomous design to reduce the personnel required on site. It isplanned that a few personnel would be required per shift for monitoring and security as the autonomous control
system will limit the need for operators.
The eVinci design has the capability for load following and grid frequency control. The eVinci autonomous
control system in combination with a micro grid, can perform load following for the most demanding situations
in remote applications. Most of the load following demands can be executed utilizing the self-control
capabilities, reactor control and power conversion control. Other high demand fluctuation will be supported
by power resources coupled to the micro grid. The eVinci system has black start-up capability supported bybatteries, for instances where the local micro grid cannot support the start-up process.

7. Instrumentation and Control Systems
The Autonomous Control System (ACS) is the primary I&C system of the eVinci Micro Reactor. The ACS
facilitates autonomous control by utilizing fiber optic sensors for high fidelity temperature measurement of thereactor, neutron flux sensors for measurement of the reactor core operation, and load following logic
programmed into the ACS. The ACS is programmed with functional logic for autonomous operation of theeVinci reactor. Temperature measurements from fiber optic sensors and neutron flux measurements from selfpowered neutron flux detectors (SPNDs) determine trip decision and load following logic. The trip decision
logic ensures the reactor does not transverse predefined operational limits. Load following logic autonomously
adjusts control drum position based on load variances, reactor temperature and neutron flux measurements.
The control drum interface logic provides a logical priority between the trip, load following logic and manualsystem controls. The ACS also collects environmental and operating data during normal operation, theautonomous logic functions control the reactor. The autonomous logic functions relinquish control of thereactor to a manual operator if required. Due to the autonomous control, no specific control room is required.

8. Plant Layout Arrangement
The eVinci Micro Reactor is designed for ease of logistics, with nearly no on-site construction and minimum
on-site installation. The eVinci system is nearly 100% fabricated in the factory and packaged in shipping
containers to be transported to site via air, water or land. After installation of the container at site, the only
connection needed between the boxes are the electrical power conduits or pipe connections if combine heatand power is required. To ensure allowable dose levels are met, integrated shielding is designed fortransportation. However, a shielding structure is planned for the reactor container to limit dose rate to operatorsand the public.
Designed for remote applications, the eVinci Micro Reactor can provide heat and power via a micro grid
system. The system can be configured for all the voltage and frequencies that might be required at remoteapplications. If power demands require several eVinci systems, they can be staged together at a site; with each
unit operating independently. Because the eVinci Micro Reactor is similar to a battery design power supply, it
does not require on-site refuelling. This eliminates additional onsite facilities and structures that would
typically be required for refuelling. After the end of its fuel cycle, the eVinci Micro Reactor will be transported
back to the factory for refuelling or for long-term storage.

9. Design and Licensing Status
The eVinci development program has completed the conceptual design phase. Westinghouse drives thematurity of the eVinci Micro Reactor product through evaluating both Technology Readiness Level (TRL) and
Manufacturing Readiness Level (MRL). The current TRL and MRL of the eVinci Micro Reactor technicalareas are estimated to be at a level 5.
Westinghouse is currently working with both the U.S. Nuclear Regulatory Commission (NRC) and theCanadian Nuclear Safety Commission (CNSC) to license the design and technology. At this point, preapplication discussions and activities have highlighted the licensing efforts to date:- Request to submit application for Vendor Design Review to Canadian Nuclear Safety Commission (CNSC)- Submitted eVinci Micro Reactor Licensing Modernization Project (LMP) Demonstration “table top” reportas part of a project to evaluate the LMP process. This report was provided to NRC for information.
- Provided the U.S. NRC a Regulatory Issue summary on process for scheduling and allocating resources forreview of a new licensing applications.

351

10. Fuel Cycle Approach
The eVinci Micro Reactor core will use either the High-Assay Low Enriched Uranium (HALEU) – Uranium
Oxycarbide (UCO) in a tristructural isotropic (TRISO) fuel or other fuel that is encapsulated.
After 3 years of full power operation without refuelling, the eVinci Micro Reactor will be disconnected andtransported back to the factory in its original canister for either refuelling and redeployment or for long-term
storage, which can be accomplished in the eVinci Micro Reactor canister it itself.

11. Development Milestones
2021
2024

352

Electric demonstration
First nuclear demonstration

ANNEX I
Summary of Global SMR Technology Development and Deployment

Figure I-1 Global Map of SMR Technology Development353

Figure I-2 General Timeline of Deployment as of 2020354

Figure I-3 Government and Private Sectors on SMR Technology Development355

Figure I-4 SMR design and technology across the world’s regions
356

Figure I-5 Stage of Design or Deployment of SMRs in terms of their Output Capacity
357

ANNEX IIPower Range of SMR Designs

Figure II-1 Power Range of Land-based Water-cooled SMRs

358

Figure II-2 Power Range of Marine-based Water-cooled SMRs

359

Figure II-3 Power Range of HTGR-type SMRs

360

Figure II-4 Power Range of Liquid metal-cooled Fast Neutron Spectrum-SMRs

361

Figure II-5 Power Range of Molten Salt-SMRs

362

Figure II-6 Power Range of Microreactor

363

364

Reactor Vessel’sheight/diameter, (m)Design status

Refuelling cycle,
monthsReactivity control

Fueltype/assemblyarray
Number of fuelassembly
Coolant
Moderator
Thermal output,
MW(t)Electrical output,
MW(e)Core inlet temp., oC
Core outlet temp., oC
Enrichment, %Reactor Type /Primary Circulation

Designorganization(s)Country of Origin

10 / 3.35

Underconstruction

11 / 3.2

Underconstruction

Control rods +Gd2O3 solid
burnable +soluble boron acid

286.5
319.5
< 4.95

284
326
3.1

Control rods

125

30

24

Light water
Light water
385

Light water
Light water100

14

UO2 pellet /17x17 square
57

Integral PWR /ForcedCirculation

CNNC (NPIC)China

ACP100IntegralPWR /NaturalCirculation
UO2 pellet /hexagonal
61

CNEA

Argentina

CAREM

Conceptual
Design15 / 5Control rods +solid burnable

24 (half core)

280
307
<5

2 x 170

Light water
Light water
2 x 540

UO2 pellet /17x17 square
76

Integral PWR /ForcedCirculation

EDF

FranceNUWARD

Conceptual
DesignControl rod +burnable absorberrods +
moderator
temperature
23 / 5

24

295.5
320.9
<5

170

Light water
Light water540

UO2 pellet /17x17 square
69

Integral PWR /Forced Circulation

KHNP&KAERI

Republic of Korea

i-SMR

Detailed
Design18.5 / 6.5

Control rods +soluble boron

30

296
322
<5

107

Light water
Light water365

UO2 pellet / 17x17
square57

Integral PWR /Forced Circulation

Republic of Korea
and Saudia Arabia
KAERI,
K.A.CARE

SMART

Detailed
Design7.9 / 4.2

Control rods +moderator
temperature

18

295
325
< 4.95470Light water
Light water1358

UO2 pellet / 17x17
square1213-loop PWR /Forced Circulation

Rolls-Royce SMR
Ltd.

UK

Rolls-Royce SMR
USA

VOYGR

Equipment
Manufacturing in
Progress

17.7 / 2.7

Control rods +soluble boron

18

249
316
≤ 4.9577 (gross)

Light water
Light water250

UO2 pellet /17x17 square
37

Integral PWR /NaturalCirculation

NuScale Power
Corporation

Table III-1: Comparison of Main Characteristics among Land-based Water-cooled SMR Designs

ANNEX III
Comparison of Main Technical Characteristics among several SMR Designs

Detailed
DesignRods + solid
burnableabsorber(B4C, Hf,
Gd2O3)
26 / 4270
288
3.81 (avg) /4.95 (max)
12 - 24

270 - 290

Light water
Light water870

UO2 pellet /10x10 array240

USA andJapan
GE Hitachi &Hitachi GE
Nuclear
EnergyBWR /NaturalCirculation

BWRX-300

365

7.2 / 2.2
Detail Design

Connected to the grid in
Pevek in December 2019.
Entered full commercial
operation in May 2020.

37
Light water
Light water
200
50
299.3
321.8
<5
30
Control rods + solid
burnable poison

China
CGNPC
loop-type PWR
UO2 pellet /17x17

ACPR50S

4.8 / 2.0

121Light water
Light water
150
35
280
316
18.6
Control rods

Number of fuel assembly
Coolant
Moderator
Thermal output, MW(t)Electrical output, MW(e)Core inlet temp., oC
Core outlet temp., oC
Enrichment, %Refuelling cycle, months
Reactivity control

Reactor Vessel’sheight/diameter, (m)Design status

Russian Federation
JSC “Afrikantov OKBM”,
PWR
UO2 pellet in silumin
matrixCountry of Origin
Design organization(s)Reactor Type
Fuel type/assembly array

KLT-40S

Basic
Design Completed

8.6 / 3.45

241Light water
Light water
175
50
282
318
< 20
Up to 120
Control rods

Russian Federation
JSC “Afrikantov OKBM”
Integral PWR
UO2 pellet / hexagonal

RITM-200M

Conceptual Design

11.2 / 2.8

52Light water
Light water
200
60
290
325
4.95
48 - 60
Control rods

Republic of Korea
KEPCO E&C
PWR
UO2 pellet /17x17

BANDI-60

Table III-2 Comparison of Main Characteristics among Marine-Based Water-cooled SMR Designs

Basic Design

57
Light water
Light water
385
125
286.5
319.5
< 4.9524
Control rods +Gd2O3 solid
burnable poison+ soluble boron
acid
10 / 3.35

China
CNNC (NPIC)
Integral PWR
UO2 pellet /17x17

ACP100S

366
490
850
14-18% LEU or
WPu
100-720 (depends
on fuel type)25

Control rods

HeliumGraphite< 600

100 - 300
587 - 633
850 - 950
14

Control rods

210

250
750
8.5

Reactor Vessel’sheight/diameter, (m)Design status

Core Discharge
Burnup (GWd/ton)Refuelling cycle,
monthsReactivity control

Coolant
Moderator
Thermal output,
MW(t)Electrical output,
MW(e)Core inlet temp., oC
Core outlet temp., oC
Enrichment, %23 / 8
Basic design

In operation

48

Onlinerefuelling
Control rods

25 / 5.7 (inner)

120

90

UO2 TRISO
ceramic coatedparticle

Preliminary Design
completed

29 / 8.2

288

HeliumGraphite
600

Coated particle fuelin compacts,
hexagonal prism
graphite blocks

TRISO
spherical
elements withcoated particle
fuelHeliumGraphite2 x 250

Russian Federation
JSC “Afrikantov
OKBM”
Modular Helium
ReactorFuel materials

Prismatic HTGR

Japan
JAEA

GT-MHR

China
INET, Tsinghua
University
Modular pebble
bed HTGRGTHTR300

Country of Origin
Designorganization(s)Reactor type

HTR-PM

Basic Design

Online fuelloading
Control rods in thereflector
15.7 / 5.6

80 - 90

250
750
10%

35

HeliumGraphite100

TRISO particles inpebbles; LEU/Th

South Africa
STL Nuclear (Pty)Ltd.
Pebble-bed HTGRHTMR100

Basic Design

16.4 / 4.88

Control rods

Online fuel loading

165

260
750
15.5

82.5

HeliumGraphite
200

UCO
TRISO/pebbles

Modular HTGR

USA
X-energy, LLC

Xe-100Preliminary Design

24 / 8.5½ core replaced every 18
monthsControl rods

325
750
14.5 (avg)
18.5 (max)
165

272

HeliumGraphite
625

UCO TRISO particle fuel inhexagonal graphite blocks

Prismatic HTGRUSA
Framatome Inc.

SC-HTGR

Table III-3 Comparison of Main Characteristics among High Temperature Gas-cooled SMR Designs

Conceptual design

12.5 / 4.6

Control rods

360

~130

550
850
~14.5 (LEU)

265

HeliumN/A
500

Modular hightemperature gas-cooled
fast reactor
UC pellet / hexagon

USA
General Atomics

EM2

367

34
N/A

61.45

36 - 78

Core Discharge Burnup(GWd/ton)
Refuelling cycle, months

24 / 3.5
Detailed design

17.5 / 26

Under construction
since July 2021

Reactor Vessel’sheight/diameter, (m)Design status

Axially movablereflectors / fixedabsorberReactivity
compensation,
emergency protection,
and automatic control
members

Reactivity control

18
Sodium
30
10
355
510
< 20

169
Lead
700
300
420
535
up to 14.5

Number of fuel assembly
Coolant
Thermal output, MW(t)Electrical output, MW(e)Core inlet temp., oC
Core outlet temp., oC
Enrichment, %U-Zr alloy

LMFR (pool type)Japan
Toshiba Energy Systems&
Solutions Corporation

Mixed uranium
plutonium nitride

LMFR

Russian Federation
NIKIET

4S

Fuel type/assembly array

Reactor TypeCountry of Origin
Design organization(s)BREST-OD-300

Detailed design forpotential
construction in 2025

8.2 / 4.53

Control rods

7 - 8 years

60

61
LBE
280
100
340
485
< 19.3

UO2 / hexagonal

LMFR

Russian Federation
JSC AKME
Engineering

SVBR

Conceptual design

6.2 / 6

Ex-core, reversedflag type B4C rods,
rotating B4C Rods

16

61
Lead
480
200
420
530
19% (avg) / 23,2%(max) in Pu
100

MOX / hexagonal

LMFR (pool type)

Italy
newcleo srl

LFR-AS-200

Conceptual design

16.7 / 7.9

Control rods

240

77

99
Sodium
286
100
355
510
13.1 (avg)

U-Zr alloy

LMFR (pool type)

CanadaARC Clean Energy

ARC-100

Conceptual
design8 – 15 years(depending on
fuel)
Control andshutdown rods;
thermal/pressure
activated,nonrodbased passive
shutdown
Approx. 9.0 / 6.7

Up to 100

USA
Westinghouse
Electric
Company,
LLC
LMFR (pooltype)
Oxide, with
provision fortransition to
Nitride
325
Lead
950
~ 450 (Net Avg.)
390
650≤ 19.75%

Westinghouse
LFR

Table III-4 Comparison of Main Characteristics among Liquid Metal-cooled Fast Neutron Spectrum SMR Designs

368
Fuel salt

Fluoride salt

Reactor Vessel’sheight/diameter,(m)
Design status

Detailed Design

Detailed design /Equipment
manufacturing inprogress

12 / 2.4

D2O level
adjustment

TRU or RGPu

<5

negativetemperature
coefficient; online
fuel addition
18.0 / 4.1

650 - 700

700

N/A

600

620

84

0.1 - 0.25

2 x 195

Refuelling
cycle, months
Reactivity
control

0.05 - 0.25

2 x 440

Thermal output,
MW(t)Electrical
output, MW(e)Core inlet temp.,oCCore outlet
temp., oC
Enrichment, %Heavy water

Graphite

Basic design
completed10.3 / 7.8

Negativetemperature coef.

12

2.3 (startup)
4.95 (makeup)

704

560

250

557

Graphite

Molten salts

UF4

7LiF-ThF -(TRU)F
4
3

Fluoride fuel salt

MSR

ThorConInternational
Thermal MSR

Copenhagen Atomic

Terrestrial Energy,
Inc.
MSR

ThorCon
USA, Indonesia

Denmark

CA Waste Burner

CanadaIMSR400

Moderator

Fueltype/assembly
array
Coolant

Country of
OriginDesignorganization(s)Reactor Type

Basic design
completed

5.4 / 5.34

Control rod, or pumpspeed,or fuel concentration

2.0(0.24%U233 +12.0%Th)
Continuous

704

565

200

450

Graphite

Molten fluoride

Molten salt with Th
and U

International Thorium
Molten-Salt Forum
MSR

Japan

FUJI

Basic Design

7 .5 / 6 (low pressure)

Strong fuel temperature coef.,
liquidneutron absorber thermometer

240

6

795

725

16

Molten eutectic fluoride
coolant salt
Solid commercial grade
graphite moderator
40

Molten salt fuel within vented
fuel tubes

MSR

Moltex Energy

UK

SSR-U
Kairos Power

USA

KP-FHR

Conceptual
Design7.2 / 3.9

Control rods,boron

Online refuelling

19.75

650

550

140

320

Li2BeF4
Fluoride salt
Graphite

Modular, pebble
bed, hightemperature, saltcooled reactor
TRISO / pebble
bed

Table III-5 Comparison of Main Characteristics among Molten Salt SMR Designs

Conceptual Design

Fuel expansion in/out
of core; Fertile fueladdition; Passive fueldraining
9.0 / 4.0

10% Pu fissile/(Pu+U
total) or ~15%
enriched HALEU
Online refuelling

750

650

50/200/400/1200

125/500/1000/3000

NaCl-XCl2-UCl3/4PuCl3-FPCly fuel salt-

Molten Chloride Salt

MSR – Fast Chloride

Elysium Industries

USA

MCSFR

369

Reactor Vessel’sheight/diameter,(m)
Design status

Thermal output,
MW(t)Electrical output,
MW(e)Core inlet temp.,oCCore outlet temp.,oC
Fueltype/assemblyarray
Enrichment, %Refuelling cycle,
monthsReactivity control

Moderator

Coolant

Country of Origin
Designorganization(s)Reactor type

FCM / Hexagonal
HALEU 19.75%
240

Control rod insertion,negativetemperature coefficient

TRISO or anotherencapsulation

5 - 19.75> 36

Ex-core control drums

Basic Design

630

N/A

Conceptual Design
Completed

300

N/A

13.25 / 3.5> 5 and > 10

2 - 3.5

N/A

15 and 30

Graphite

Helium

USA
Ultra Safe Nuclear
Corporation
HTGR

MMRTM

7 - 12

Metal hydride

Heat pipes

USA
Westinghouse ElectricCompany LLC
Heat pipe cooled Reactor

eVinciTM

Detailed Design

5 (reactor fits 12.19 m
long ISO container)

Redundant independentbanks of Control
drums & Shutdown rods

19.95
96

TRISO-UCO /Hexagonal

855

590

10

22

Graphite

Equipment manufacturing inprogress

2.55 / 1.22

> 60

19.75
Up to 20 years

Uranium Zirconium Hydride

548

480

0.015 - 0.027

0.075 - 0.1

Hydrogen in fuel

NaK eutectic

Liquid Metal Cooled Thermal
reactor

HTGR
HeliumUSA
Idaho National Laboratory

MARVELUSA
HolosGenHOLOS-QUAD

Table III-6 Comparison of Main Characteristics among Microreactor Designs

Conceptual Design

2.5 / 6.0

In-Ga Expansion
Module4.8 - 5.0
Continuous

Silicide (U3Si2) /Hexagonal

685

680

3-4

Heat-Pipe cooled andcalcium-hydride
moderated reactor
Sodium heat-pipecooledCalcium hydride(CaH2)
10

Japan
Toshiba Corporation

MoveluX™

AMR

Pre-conceptual
designControl andshutdown rods in
core, and thereflector
5 .96 / 2.78

T RISO particles/
LBE eutectic/ SiC
tubes
10 - 20
96

750

450

3

10

GraphiteSouth Africa
STL Nuclear (Pty)Ltd
HTGR(advanced
prismatic)
HeliumANNEX IV
Non-Electric Applications using Small Modular Reactors
The interest in non-electric applications of nuclear power is growing, driven by a series of factors,
including environment, economics, and security of energy supply.
There is a massive need for low carbon heat and low carbon fuels, as well as the technologies to produce
them using nuclear power, such as SMRs, in order to support the decarbonization of the ‘hard-to-abate
sectors’ i.e. those that cannot be electrified, such as metal refining, synthesis of chemical feedstock,cement and steel production, and heavy duty transport. In particular, nuclear power technologies canprovide high-temperature heat needed by industrial processes. One of the low carbon options to beconsidered to achieve this is the capitalization of the nuclear heat provided by SMRs.
The emerging SMRs and their use for non-electric applications come with additional challenges, as
their deployment requires a ‘level playing field’ in terms of policies, support to innovation, financing,as well as clear market signals to favour low carbon technologies and the associated licensing process.
SMR concepts being currently developed worldwide are of major reactor technologies that can support
a broad range of applications in addition to electricity generation, including applications in the industry.
SMRs bring the possibility and the opportunity to have their design optimized for non-electric
applications. This involves single or multi-purpose use – generating commodities like hydrogen or
potable water, providing heat for industrial applications and district heating, together with or without
electricity generation for the grid.
Nuclear cogeneration is the generation of electricity and useful heat jointly that applies to nuclear powerreactors in general, including SMRs. The followings are certain benefits that come from considering
nuclear cogeneration:Enhanced efficiency: the efficiency of a nuclear power plant operating in cogeneration mode canexceed 80%, compared to 33% in the case of traditional large light water reactors operating justfor electricity generation. It also offers additional uses of nuclear energy by capitalizing on heatfor district heating, seawater desalination, hydrogen production or process heat for industrial
applications.



Optimum use of energy by avoiding unwanted transformations of energy.

Improved flexibility, allowing switching between the two outputs – electricity and heat.



Cogeneration gives nuclear power the flexibility to function in an energy system where a growing
proportion of electricity comes from intermittent renewables, allowing plants to switch fromelectricity generation to other applications when demand is being met by wind, solar, or otherenergy sources.



Reduced environmental impact: less heat waste is discharged in the environment; less water isneeded for cooling; and using of nuclear heat can eliminate the need for fossil fuels for heat.

On the type of technologies, light water reactors are best suited to district heating and desalination,owing to their low working temperature range. While other technologies with higher-working
temperatures, such as liquid metal cooled reactors (550°C), molten salt reactors (800°C), hightemperature gas-cooled reactors (850°C), and very-high temperature reactors (900°C to 1000°C) are
better suited for industrial processes. Among the industries requiring high temperatures there arepetroleum refining, coal gasification, steel making, and hydrogen production using high-temperatureelectrolysis, steam methane reforming or thermochemical processes. Furthermore, these nuclear
reactors can be used for desalination and district heating when used as cogeneration systems. The figure
below presents an overview of these applications together with illustrative SMR designs.

370

District heating

A heat distribution network would require steam or hot water of 80 - 150°C that can be supplied
by a nuclear reactor, with a typical pipeline distribution range of 10 - 15 km. There are well provenprojects for using nuclear combined heat and power for district heating using large scale
nuclear reactors in several Member States. However, the market is still limited and there are
deployment issues for this low share, such as the long distance between nuclear sites and urban
areas, local governance, economic feasibility, institutional structures, and the difference ofcharacteristics of national energy systems. These challenges associated with using largecommercial reactors for district heating are now being addressed by SMR designs developed
particularly for this purpose, for example in China and Finland.
Water desalination
Nuclear desalination has been demonstrated as being technically sound and economically affordable.
There is accumulated experience worldwide in using large-scale nuclear reactors coupled to water
desalination plants. Water desalination is an energy-intensive process. It can be achieved throughthermal desalination processes - requiring heat to separate the distillate from high salinity water, reverse
osmosis that uses membranes and high pressure to separate salts from the water, or hybrid processes,
requiring both electricity and heat. The temperature range required is 70 - 110°C. Today, a series offactors, such as environment, energy reliability, energy supply security, the need of potable water anda significant growth in energy demand, are driving a high interest in the development and expansion of
nuclear energy options for this application, where SMRs are showing high potential due to theirversatility in use.
Industrial heat applications
Some of heat-intensive industrial processes can benefit from nuclear heat, including enhancing brown
coal quality, coal liquefaction, coal gasification, and enhancing oil recovery. The main requirements
are that the cogeneration nuclear plant has to be located close to the industrial user and to have a highreliability that nuclear can provide. This makes SMRs particularly attractive for these applications. The
transport of the heat to the end user can become a challenge as the longer the distances the higher the
associated costs with distribution and transport. This is the reason for which in the case of using theheat from nuclear reactors, the end user should be co-located or located close enough to the nuclear
power plant. SMRs allow co-location with industrial end users due to the potential to adopt reduced
emergency planning zone size, providing that there is acceptance of it by nuclear regulators. There has
to be also a governance in place between industrial users and the utilities operating the nuclear reactor
that supplies the process heat.
Hydrogen production
Nuclear energy can support the increasing needs for hydrogen production, through providing clean
electricity, as well as heat. The IEA report on Nuclear Power and Secure Energy Transitions(International Energy Agency, Nuclear Power and Secure Energy Transitions. From today’s challenges
to tomorrow’s clean energy systems, June 2022) released in June 2022 indicates that the economics forhydrogen production would be more favourable if the nuclear reactor is co-located with the hydrogenusers, hence minimizing transportation costs. SMRs are well positioned to support production ofhydrogen in the vicinity of end users.
While nuclear cogeneration and non-electric applications can be supported by the different reactor
technologies of different sizes, their attractiveness and applicability may increase when it comes to
SMRs, based on the following aspects:


The nature of industrial heat market is highly fragmented, with the majority of potential heatneed included in the 1 - 300 MW(t) range and 50% requiring less than 10 MW(t). This would
bring an additional challenge for large scale nuclear reactors as the economics of using large371

capacities to capitalize on a small amount of heat needed is not favourable; instead, SMR
designs are more suitable to support industrial heat applications.
Moreover, nuclear heat used for industrial applications has to have a reliable and available
source, and the source has to provide maximum availability. The SMR technologies arebringing the possibility of less frequent refuelling and can be designed from the beginning as
multipurpose or to support solely producing a commodity (hydrogen or potable water).

The SMR thermal ratings are about one-third or less of the large existing reactors. This helps
lower investment requirement for the deployment and thus greatly improve the affordability tomore countries, private investors, and industrial users. The modularity also makes them readily
scalable with the scale and increment of energy demand by most industrial heat process plants.



SMRs bring a relevant technology for load following and the overall power at the site level canbe fractioned, facilitating an optimal harnessing of the energy produced and allowing it towards
electricity production, conversion into commodities or storing for a later use. Nuclear power
plants composed by multiple small modular reactors have as key advantage over a single
equivalent large reactor (in terms of capacity) the possibility to operate some of the SMR fleet
at full capacity while directing the thermal power of some modules for non-electric applications
only.



SMR designs allow for optimization of the use of reactors operating in cogeneration mode andgenerating multiproduct, the optimization being based on criteria such as cost of products,
return of investment and energy efficiency.

Coupling nuclear reactors with non-electric applications can also provide energy system storage– i.e., storing energy in the form of heat or as an energy vector such as hydrogen. This is thebasic concept of hybrid energy systems.

SMRs are flexible systems that can be adapted for non-electric applications, according to their key
operating parameters, i.e., exit working fluid temperature. They can be integrated effectively into theenergy systems supporting communities from largest cities to most remote locations, bringing in theopportunity to support applications beyond solely electricity generation.

372

373

Figure IV-1 Exit Working Temperature of SMR Technologies and the Corresponding Non-Electric Applications

ANNEX V
Economic Challenges in Deployment of Small Modular Reactors

Introduction
SMRs have unique design, safety and economic features that make them attractive to potential project
developers and end-users but also investors, governments, and communities worldwide. The technology
can be applied to produce low-carbon electricity, heat (district heating, for example), hydrogen anddesalinated water, in addition to grid, load-following and ancillary services. More than 80 SMR designs
and concepts are currently under development and have varying degrees of readiness levels. For eachof these projects, development costs need to be understood, as well as construction and operation
expenses, which still need to be appropriately estimated, analysed and optimised. Specific revenue
models are also needed for demonstrating the business case and secure access to funding, financing,and low cost of capital for the promoters of the technology. Finally, the macroeconomic impact
associated with SMR design development, manufacturing, construction and operation (including
periodical maintenance) has to be quantified and communicated to gain the support of the governmentand society at large.

Figure V-1: Economic challenges facing SMR
This annexe provides an overview of the economic challenges facing SMR during the key phases of atypical project (Figure V-1) – from the project development phase to construction and commissioning
to the operation phase – with a focus on costs and cost drivers, funding and financing issues, andeconomic impact.
Costs and cost drivers
Power generation systems, including SMR, are usually characterised by the metric of Levelized Costof Energy (LCOE). LCOE represents the per-MWh cost (in discounted real dollars) of building andoperating a generating plant over an assumed economic lifetime. The key economic indicator isdetermined by setting to zero the net present value (NPV) of cash in- (from the sale of electricity and,possibly, other products and services) and out-flows (CAPEX and OPEX) incurred over the key phasesof a power generation project (Figure V-2).

374

Figure V-2: LCOE estimation basis
Constructing NPPs usually involves high capital costs with a long construction period, followed by anoperating phase with relatively low and stable recurring expenses, translating into an LCOE primarily
driven by capital expenditures and the discount rate [1].
Another key driver is the amount of generated power, which is directly linked to the capacity factor. By
producing multiple by-products and services (a combination of electricity, heat and flexibility), SMRs
can maximize the utilization of the generation asset over its lifetime and secure earnings, including
when operating in a load-following mode.
SMRs are based on the repeated manufacture of a modular and standard design of components andsystems in a factory setting, potentially leading to a decrease in capital expenditures, a shortening of
the construction and commissioning period, and a low cost of capital. The in-factory fabrication model
is widely applied in the airline and shipbuilding industries and is a synonym for learning, cost reduction
and improved quality [2]. To replicate this model, SMR developers need to embrace manufacturing bestpractices from other industries while sustaining demand for standardised SMR components and systems
to enable learning and diffusion of know-how and achieve significant cost reductions. They also needto deal with a regulator focusing primarily on safety, and not necessarily on the speed of delivery [3].
A study commissioned by the Department of Energy and Climate Change of the UK suggests that SMRs
could achieve LCOE parity with large units at 5 GW of cumulative deployment, assuming a production
rate of 10 SMR units per year [4].
Funding and financing issues
Nuclear power projects have particular characteristics and risk profiles, which makes it difficult for
their promoters to secure financing and benefit from a low cost of capital:Implementing new nuclear power plant projects involves a long, complex and contextdependent regulatory process.
Nuclear projects are also subject to country-specific risks, including political risks, a major
concern for investors and lenders that are not easily assessed or mitigated.

SMRs share many of the attributes of GW-scale reactors, including exposure to nuclear- and countryspecific risks (Figure V-3), making them less attractive to risk-averse debt and equity providers thanrenewable generators. However, SMRs are smaller, less complex, easier and faster to build andcommission, which could translate into improved access to financing. Furthermore, the modular

375

character associated with the technology brings some flexibility, e.g., in deciding the timing and themagnitude of an increment in the installed capacity, which can make the technology more appealing toinvestors. Finally, SMRs have the potential to produce a combination of electricity, heat and flexibility,
diversifying its sources of income and mitigating market risks, which can also improve its
attractiveness.

Figure V-3: Risks: Internal vs external
Enabling the production and commercialisation of SMRs will require developing and pursuing derisking strategies to channel funding to the project and keep costs under control while securing revenues
for nuclear power plant operators. Another key enabler is a policy framework recognising the value
nuclear can bring to an evolving energy system, low-carbon, highly coupled, and probably renewablesdominated. Governments have a crucial role in creating such an enabling environment and overcoming
obstacles to introducing SMRs and other low-emitting energy sources.
As for large units, SMRs can be funded by the State (public financing). Alternatively, a utility can build,own and operate the infrastructure, borrowing funds, raising equity capital, or simply relying on its own
balance sheet (corporate financing). Project financing (recourse financing) can also be an option on thecondition that the government accepts to cover nuclear- and country-specific risks. Finally, projects
conducted overseas can benefit from the involvement of Export Credit Agencies (ECAs), which tend
to increase the lenders' confidence.
Revenues can be secured through various models, e.g., the Contract for Difference (CfD) [5] andRegulated Asset Base (RAB) [6] models in the UK, depending on the market's value rewarding
mechanisms, i.e., how electricity, heat and ancillary services are sold and priced. Today's markets
primarily reward output, i.e., the MWh of energy produced, which can be traded on the futures or spot
markets. The provision of capacity is also rewarded through the procurement (by the transmission
system operator) of balancing services but to a much lesser extent.
Economic impacts
Nuclear power projects are often described as labour intensive, paying higher wages compared to othertechnologies, and having long-lasting economic impacts through direct, indirect and induced effects(Figure V-4), which can be estimated and quantified [7]. This impact was historically significant incountries with a thriving nuclear industry and a high degree of localisation, as emphasised by a joint
report of the NEA and the IAEA that studied the employment effects of nuclear energy programs [8]and a recent study on "multipliers" of the International Monetary Fund (IMF) [9]. As for the large
reactors built in the 1970s, 1980s and 1990s, SMR projects are expected to stimulate manufacturing
and engineering services in all countries participating to the supply chain, as well as construction,
operation and maintenance activities in the end-user country, impacting the labour market andgenerating economic growth across a wide range of economic sectors beyond these perimeters.

376

Figure V-4: Macroeconomic impacts: direct, indirect and induced effects
SMR projects can also potentially impact countries adding SMRs to their energy mixes, even in theabsence of manufacturing and construction activities. By providing electricity at an affordable - andpredictable - price, SMRs can, in fact, trigger induced effects within the income-consumption loop
(Figure V-4), resulting in benefits to a variety of actors in the economy.
Table V-1 provides an example of an economic impact assessment associated with developing anddeploying five grid-scale 300 MWe SMR projects in Canada [10]. In addition to job creation figures,the study, prepared at the request of the provinces of New Brunswick, Ontario, and Saskatchewan,
reports a positive impact on GDP of 17 billion dollars; and an increase in government revenues
amounting to 5.4 billion dollars.
Phase

Employment

Project dev.

1,528

Construction

12,455

Operation

1,469

Decommissioning

1,193

Table V-1: Direct, indirect, and induced employment on an average annual basis [10]377ANNEX VI
Fuel Cycle Approach adopted in SMRs
Introduction
Design and operation of SMRs are building on experience gained from current power and research
reactors in assessing and planning for spent fuel management, waste management, anddecommissioning of SMRs. For SMRs to be sustainable, it is necessary to plan for decommissioning
and management of spent fuel and waste early in their development; this ensures that the rightinfrastructure exists for waste and spent fuel management, facilitates confidence of interested parties,
and avoids unwanted liabilities and legacies for future generations.
To the extent that certain SMR designs concepts, associated fuel cycle decisions and future operational
procedures, are similar to those used in current operating large NPPs, the spent fuel management (SFM)
and radioactive waste management (RWM) activities can most probably be implemented consistently.
Advanced SMR designs will generate entirely new wastes streams and new spent fuels which must be
handled safely. Early planning will help avoid any unforeseen technological, environmental, or
financial difficulties.
A better understanding of the specific fuel cycle approach being proposed by the various SMR designs
is very important to stakeholders in Member States. Particularly, the knowledge of the fuel cycle will
facilitate discussion on the front-end requirements and fuel supply, the potential management routes forspent fuel and waste forms (discussed in the annex VII), the potential needs for a final waste repository,
and Safeguards-by-Design implementation by SMR designers. It should be recognised that fuel cycle
choices are often made by the state and this may place restrictions on vendors that wish to deploy an
SMR in a country.
There are many designs being advanced from major technology lines to meet the development
objectives of SMRs. SMRs are not only of water-cooled reactor types but include many designs utilisingother coolant types or neutron spectra, including HTGRs, LMFRs and MSRs represented in this booklet.
Many of these designs adopt an innovative approach for their fuel cycles. Most of the SMR designs
propose longer operation cycles between refuelling. This might raise issues on fuel (and cladding whereapplicable) performance (e.g. at higher burnup), and other aspects such a core maintainability, on-line
access to service key components and the different steps for managing the spent fuel (mainly storage,transport and recycling routes). Some of the new fuel cycles and new fuels proposed include the use of
Th/233U, reprocessed uranium (RepU), MOX, transuranic (TRU) fuels through the reprocessing of spent
LWR fuel. To achieve that it is necessary to include advanced technologies as pyro-processing for SMR
fuel reprocessing and recycling. It is important to highlight that some of those innovative routes are notindustrially implemented yet, a few are at demonstration stage and most of them are at R&D stage, witha low technology readiness level (TRL).
Land-based water-cooled SMRs
For water-cooled SMRs, various fuel cycles are adopted depending on the design approach andobjectives. Land-based water-cooled SMRs typically utilise existing LWR fuel design with 235U
enrichment below 5% to achieve a range of discharged burnups as low as 30 GWd/tHM for district
heating reactors, and between 2.3 up to 162.4 GWd/tHM for electricity generation, to attain refuelling
cycles of 18, 24, 36 or even up to 96 months. The ultimate objective is to have an optimized capacity
factor of higher than 90% while minimizing the operation and maintenance (O&M) costs. Longer fuelcycles improve the overall plant availability but could be at the expense of a somewhat reduced
discharge burnup.
Typical PWR-type SMR use fuel with enrichment lower than 4.95%, propose a refuelling cycle of 1824 months and replace half of the assemblies every cycle to optimize the overall fuel economics while
maximizing the discharge burnup. A 3-batch refuelling design, on the other hand, improves the378

discharge burnup and fuel utilization. The target refuelling outage is between 18 to 36 days. A natural
circulation integral PWR design for a multi-unit plant, for example, adopts a 24-month refuelling cycle
and a 3-batch in-out refuelling scheme. This means that during the refuelling process, one-third of thefuel assemblies are removed from a reactor module and placed in the spent fuel pool and one-third offresh assemblies are introduced; while at the same time the remaining assemblies are re-shuffled in thecore.
The selected option for the standard design of CANDU SMR is an open fuel cycle based on naturaluranium, producing a very low residual uranium spent fuel with low heat generation. It is also designed
to be able to burn RepU from LWRs, making it a valuable addition to any light-water fleet. The use of
MOX and Thorium fuels is also possible with proper customization.
There are also converted PWR type SMR designs originally deployed for marine-based SMRs. Thesedesigns have enrichment close to the 20% upper limit of LEU to ensure long time operation in remoteareas while still satisfying international proliferation goals. The target refuelling cycle is as long as 72– 84 months. In the framework of transportable SMR designs, a factory-fabricated reactor vessel could
be delivered to the site loaded with fresh fuel. This initial load is designed to provide the whole NPP
lifetime without refuelling.
This booklet also reports BWR-type SMRs with natural circulation. During the refuelling outage 15%
to 25% of the bundles in the core are replaced by fresh fuel with the result that they stay in the core forseveral cycles, until they are discharged. When removed from the core, spent fuel is stored in the fuelpool inside the reactor building for 6 to 8 years then it is transferred to storage casks that are stored ina facility located outside the reactor building, waiting to be disposed of as typically, for the standard
BWR designs an open cycle is used for managing the spent fuel.
In general, the water-cooled SMRs provide a spent fuel pool using storage racks with storage capacityfor 6 to 10 years after operation . A longer storage option of up to 20 – 30 years could be proposed forsome designs, depending on owner’s requirements. The spent fuel pools are connected to the ultimate
heat sink through dedicated engineered safety features. The spent fuel storage racks include not only
sufficient storage for many years of operation, but also to store potential defective fuel assemblies andfor non-fuel core components such as control rod assemblies.
Ultimately, the fuel cycle can be tailored to customers’ requirements in compliance with the regulatory
requirements both in the countries of origin and host countries.
Marine-based water-cooled SMRs
The first SMR design that completed construction, was connected to the grid and is now in commercial
operation is the PWR-type KLT-40S installed in the Akademik Lomonosov floating power unit (FPU),
deployed in Pevek, Russian Federation. The purpose of the FPU is to supply cogeneration of electricity
and process heat in difficult-to-access remote regions. Therefore, the adopted fuel cycle has to becommensurate to the purpose of deployment. In Russia, there are at least five (5) marine-based SMR
designs (included in this booklet) and their variants. To achieve a long refuelling cycle of 30 – 36
months, a near 19% fuel enrichment is required and 45.4 GWd/ton core discharge burnup is achievedin KLT-40S. Refuelling is performed 14 days after reactor shutdown when the levels of residual heatreleases from spent fuel assemblies have come down to a required level. No special maintenance or
refuelling ships are necessary. A single batch fuel loading is done to attain a maximum operation period
between refuelling. In other words, fresh fuel is loaded in all the core positions replacing the burned
fuel assemblies.
The RITM-200M based optimized floating power unit (OFPU) is also designed as a transportable SMR
power plant that has a refuelling cycle up to 120 months. The OFPU is delivered to the site with fresh
fuel in its reactors. After completion of the fuel cycle, the OFPU returns to the country of origin together
with the spent fuel in its reactor. All operations for production, post-reactor maintenance andreprocessing of spent nuclear fuel are performed in the country of origin.

379

HTGR-type SMRs
The objective of HTGR-type SMR designs is built on inherent safety characteristics with the focus on
accident-tolerant fuel (ATF). The HTGR-type SMRs come as pebble-bed or prismatic core designs
adopting tristructural isotropic (TRISO) coated particle fuel or as designs with hexagonal assemblies
adopting UC or UO2 fuel. The SMR of this type in operation is in China and was connected to the gridin December 2021. It is a 2-unit pebble bed type HTR-PM that can generate 210 MW(e). Online
refuelling (pebble bed) or long refuelling cycles (prismatic) of 25 to 60 months or even no refuelling(no shuffling) are adopted by near-term deployable HTGR-type SMRs with the 235U enrichment ranging
from 8.5% to 19.75%. The typical discharge burnup of the HTGR-type SMRs is substantially higher
than that of typical water-cooled SMRs. It ranges between 60 to 165 GWd/tHM depending on fuel typeand fuel cycle. The higher enrichment is needed for the increased burnup target but also to compensate
for the typically larger neutron leakage from this graphite moderated and reflected core with its lowpower density but also relatively small active core diameters (needed to limit fuel accident
temperatures). Although their diameters are physically larger than typical LWR SMRs, the cores areneutronically much smaller leading to the aforementioned higher net leakage.
The EM2 high-temperature gas-cooled fast reactor uses an open fuel cycle with LEU/DU vented fuelsthat can exceed 30 years operation without refuelling or shuffling. This should lead to reduced cost anddecreased proliferation risk, while achieving high fuel utilisation with low mass of waste streams. The
core is capable of burning used LWR, plutonium, and thorium fuels. The spent fuels are stored in storage
facility at the reactor site. After cooling, the spent fuel will be directly disposed of or recycled. Whenthe closed fuel cycle scheme is applied, the spent fuel would undergo pyroprocessing for metallic fuels
and would be refabricated as fresh fuel for reuse. There is no reprocessing plant based on pyro
technology under operation so far. This technology is still under demonstration stage with low TRLs
and technical issues to be solved as materials corrosion, etc.
Both open and close fuel cycles can be adopted with options including Th/U, Th/Pu, Pu, and actinide
fuel forms . Most of these fuels have been developed, manufactured and demonstrated (or at least tested)in the past HTGR projects. In HTR-PM, for example, if recycling is adopted, the spent fuel spheres
would be dismantled, and the nuclear fuel would be reprocessed (in normal reprocessing facilities, that
may have to be adjusted to receive the material). Currently there are limited routes available or decided
to manage graphite wastes.
Most HTGR-type SMRs will start with open cycle option and no reprocessing will be done. The
generated spent fuel shows some of the favourable proliferation resistance characteristics of thesereactors, one being that the total plutonium and 239Pu assay (per unit of energy produced) is much lower
due to better in-situ utilisation. In the case of pebble bed reactors, fuel that has not reached the target
discharge burnup is recycled again through the reactor and remains in the closed safeguarded fuelhandling system until it is classified as spent fuel; and then loaded into a spent fuel cask.
With higher thermal efficiency and high fuel burnup, HTGR-type SMRs support sustainability for open
fuel cycles. Most of the core designs are also compatible with various more advanced fuel cycles
employing fertile / fissile material conversion and recycle including Th/U, Th/Pu, Pu, and actinide fuelforms. The TRISO coated particle fuel could also be recycled when and if such process is mandated
and economically viable, although the reprocessing routes are still at R&D and demonstration stages.
Liquid metal-cooled fast neutron spectrum SMRs
This booklet reports eight liquid-metal cooled fast neutron spectrum SMRs. In general, compared towater-cooled SMRs, LMFR-type SMRs have a higher enrichment of ranging from 14% to slightly
less than 20% 235U (some fuels also with plutonium loadings) with core discharge burnup in the rangeof 60 to 100 GWd/tHM to realize longer fuel cycles of up to 30 years.
Within this category, BREST-OD-300 is a lead-cooled fast neutron reactor facility that has started
construction in June 2021 in Seversk, Russia with completion schedule by 2026. The project is to enable380

a closed nuclear fuel cycle (CNFC) for full utilisation of the energy potential of natural uranium. Mixed
nitride fuel with high density and thermal conductivity ensures optimum performance of fuel in the core
(core reproduction ratio ~ 1.05) to compensate for burnup. The fuel type considered for the first core
and the first partial fuel reloads of the BREST-OD-300 fast reactor is nitride of depleted uranium mixed
with plutonium, whose composition corresponds to that of irradiated spent fuel from VVER’s following
reprocessing and subsequent cooling for about 25 years. After completion of the initial stage the reactor
will operate a closed nuclear fuel cycle. For the production of new fuel, it uses the products from thereprocessing of its own spent fuel.
The LFR-AS-200, with MOX fuel of 19% avg / 23.2% max in Pu and burnup of 100 GWd/ton, can benearly self-sufficient in Pu with a breeding ration of ~ 0.9. With some modifications, it can also betransformed into Pu burner with a breeding ration of ~ 0.5.
The Westinghouse Lead Fast Reactor, with fuel enrichment up to 19.75% and core discharge burnup
up to 100 GWd/ton, applies once-through fuel cycle with single-batch long-life core design. However
close fuel cycle is a possible option.
Another LMFR-type SMR design – SEALER-55, with fuel enrichment of 12%, core discharge
burnup of 60 GWd/ton and fuel cycle approach to be decided, adopts uranium nitrid fuel which
wille be fabricated by direct ammonolysis of 12% enriched UF6, followed by spark plasma sinteringof pellets.
Molten salt SMRs
Molten salt reactors (MSRs) have different characteristics from that of solid-fuelled reactors. In
MSRs, as no fuel structure or cladding is required, the fuel is not subject to failures due to highburnup or mechanical damage. The fuel is already in a molten state so there is no risk of fuel melting(with the severe consequences that may challenge the vessel integrity and possible fission
products release). Molten-salt reactors reported in this booklet adopts different fuel cycles that aim
for longer fuel cycle up to 150 months, online refuelling (adding of fuel containing molten salt) andcontinuous operation. Enrichment levels vary from those designs with less than 5% enrichment tosome with a higher-level enrichment up to 19.9%.
In the IMSR400, which plans to complete vendor design review (VDR) Phase 2 in 2022 in Canada,
the core-unit will be replaced at the end of its 7-year fuel cycle. The de-fuelled core-unit is thenmoved by overhead crane from its operating silo to a long-term storage silo inside the reactor
auxiliary building. A third core-unit can then subsequently be installed in its place ready to begin
operations when needed (or once the second operating core-unit reaches its end-of-cycle and is shutdown).
The Copenhagen Atomics Waste Burner uses LiF-ThF4-(TRU)F3 or LiF-ThF4-RGPuF3 kickstarter
fuel salt and LiF-ThF4 blanket salt, where the 233U production benefits from the high number ofexcess neutrons from the plutonium fissioning. As the thorium fuel cycle converges towards
equilibrium the breeding process benefits from 233U superior neutron economy.
Another MSR design is ThorCon. It adopts enriched uranium of 2.3% as startup and eventually
4.95% after makeup, with discharge burnup of up to 145.8 GWd/tHM. However, once HALEU
becomes available in power plant level quantity ThorCon with 500 MWe can operate as a thorium
converter. The ThorCon core will require 5.3 kg of 19.7% enriched uranium and 9.0 kg of thorium
per day to be added to the core to realize the 8-year core life. During the 8-year fuel cycle, a portionof the fertile thorium is converted to fissile 233U which then becomes part of the fuel.
A potential issue on the fuel cycle of MSR is the management of the spent molten salt. The complex
mixture of fuel and molten salt which will be difficult to manage as a high-level waste. The potential
reprocessing routes require pyro technologies and there is no pyro-reprocessing plant under operation
so far.

381

Microreactors
The microreactors reported in this booklet are of different technology lines: pressurized water reactor,
heat pipe-cooled reactor, high temperature gas-cooled reactor, molten salt reactor, and liquid metalcooled fast neutron spectrum reactor. The range of electrical output is from dozens of kilowatts up to
30 MW(e). The adopted fuel enrichment is between 4.8% up to 19.95%, with continuous online
refuelling or long refuelling cycle between 36 months up to 20 years. Limited information of the fuelcycle aspects of microreactors are available.
The Westinghouse eVinci uses heat pipes for heat transfer and HALEU Uranium Oxycarbide (UCO) in
TRISO encapsulated fuel as one of the options that can be used. After 3 years’ operation without
refuelling, the microreactor will be disconnected and transported back to the factory in its original
canister for either refuelling and redeployment or for storage waiting for a disposal route.
HOLOS -QUAD also uses TRISO-UCO fuel. Its spent fuel cartridges passively dissipate decay heatfrom reactor shutdown, and can be placed within dual purpose transport and storage shielding dry casks
within 72 hours from shutdown. The spent fuel cartridge dimensions altogether with those of the drycask are designed to comply with ISO shipping container dimensional constraints. The spent fuelcartridges will be stored at long-term storage facilities and the spent fuel in them will not be processedas the type of fuel utilized is already in a stable form for long-term storage.
Another product called the Micro Modular Reactor (MMR) uses Fully Ceramic Micro Encapsulated(FCM™) fuel which is based on very small particles containing low-enriched uranium. The TRISO
fuel, which contains the radioactive by-products of fission within layered ceramic coatings, is further
encased within a fully dense silicon carbide matrix. This is like encasing the fuel in a diamond-like
substance. The combination of TRISO fuel particles and the silicon carbide coating provides anextremely rugged and stable fuel with extraordinary high temperature stability. FCM fuel ensurescontainment of radioactivity during operations and accident conditions, which means that no fission
products are released from the fuel. The MMR is initially fuelled for a 20-year cycle, and may be
refuelled for an additional 20-year cycle. There is no long-term fuel storage on site.

382

383

Open Fuel Cycle

Enrichment <5%

LongerRefuelling Cycle(≥ 24 months)

Closed Fuel
Cycle

CAREM, CANDU SMRTM, HAPPY200, iSMR, SMART, VK-300, RUTA-70, STAR,
Rolls-Royce SMR, VOYGRTM, BWRX-300,
Westinghouse SMR, mPower, BANDI-60,
SHELF-M

ACP100, CAP200, NHR200-II, NUWARD,
IMR, i-SMR, SMART, RITM-200N, VK-300,
KARAT-45, KARAT-100, RUTA-70, STAR,
SMR-160, OPEN20, KLT-40S, ACPR50S,
ACP100S, BANDI-60, ABV-6E, RITM200M, VBER-300, SHELF-M
CAREM, ACP100, CANDU SMRTM,
CAP200, DHR400, HAPPY200, NHR200-II,
TEPLATORTM, NUWARD, IMR, SMART,
VK-300, KARAT-45, KARAT-100, RUTA-70,
Rolls-Royce SMR, VOYGRTM, BWRX-300,
SMR-160, Westinghouse SMR, mPower,
OPEN20, ACPR50S, ACP100S, BANDI-60

TEPLATORTM, NUWARD, IMR, RITM200N, KARAT-45, KARAT-100, RITM-200M

Water-cooled Reactors

Fuel Cycle
Requirements /Approach

STARCORE, GTHTR300, GTMHR, FMR

HTR-PM (currently),
STARCORE, GTHTR300, GTMHR, MHR-T, MHR-100,
AHTR-100, HTMR100, EM2,
FMR, Xe-100, SC-HTGR,PeLUIt, HTR-10, HTTRHTR-PM (future), JIMMY

BREST-300-OD, ARC100, SVBR, SEALER-55,
Westinghouse Lead Fast
Reactor

BREST-300-OD, ARC-100(to be considered when
reprocessing is licensed
and accepted), 4S(applicable),
MicroURANUS, LFR-AS200, SVBR (in moredistant future), SEALER55 (optional),
Westinghouse Lead Fast
Reactor (possible)

ARC-100, 4S (applicable),
SVBR (on the first stage),
SEALER-55 (optional),
Westinghouse Lead Fast
Reactor,IMSR400, SSR-W,
Copenhagen Atomics
Waste Burner (optional),
ThorConIMSR400, smTMSR400, THORIZON, SSRU

smTMSR-400, CMSR
(offsite reprocessing),
FUJI, THORIZON,MCSFR, LFTRIMSR400, SSR-W,
SSR-U, KP-FHR, Mk1
PB-FHR

SMR designs by type of coolants and technology characteristics
Liquid-metal cooled
Molten Salt
HTGRs
Fast Reactors
Reactors

Table VI-1 Fuel Cycle Approach adopted by SMR Designs

MoveluX

Energy Well, AMR

MoveluX (optional),
LFR-TL-30

Energy Well,
MoveluX, HOLOSQUAD

Microreactors

384

Use of Spent Fuelas FuelUse of Thoriumcycle and/or
Disposition of
Plutonium

Enrichment >
20%

5% ≤
Enrichment ≤20%

TEPLATORTM

KLT-40S, RITM, SHELF,
ABV-6M, STARCANDU SMRTM (possible with suitablecustomization)

RITM-200N, STAR, KLT-40S, ABV-6E,
RITM-200M, SHELF-M

STARCORE (plan to investigate),
GTHTR300 (applicable), MHR100 (possible), AHTR-100(flexible), PBMR®-400 (flexible),
HTMR100 (optional), EM2(capable)
GTHTR300 (applicable), EM2(feasible)

HTR-PM, STARCORE, JIMMY,GTHTR300, GT-MHR, MHR-T,
AHTR-100, PBMR®-400,
HTMR100, EM2, FMR, Xe-100,
SC-HTGR, PeLUIt, HTR-10,
HTTR

BREST-300-OD (DU
mixed with plutonium),
MicroURANUS, SVBR

LFR-AS-200 (19% avg /23.2% max)
LFR-AS-200 (capable),
SVBR

BREST-300-OD, ARC100, 4S, MicroURANUS,
SVBR, SEALER-55,
Westinghouse Lead Fast
ReactorsmTMSR-400,
Copenhagen Atomics
Waste Burner (possible),
SSR-U, Mk1 PB-FHR(possible), MCSFR,
LFTR
SSR-W, smTMSR-400,
Copenhagen Atomics
Waste Burner, FUJI,
MCSFR

smTMSR-400, CMSR,
FUJI, SSR-U, KP-FHR,
Mk1 PB-FHR, MCSFR,
ThorCon (possible)

AMR (to beinvestigated), LFRTL-30

Energy Well, ELENA,
UNITHERM, AMR,
LFR-TL-30, UBattery, HOLOSQUAD, MARVELANNEX VII
Spent Fuel, Waste Management and Disposal Plans adopted for SMRs
Introduction
Design and operation of SMRs are building on experience gained from current power and research
reactors in assessing and planning for spent fuel management, waste management, and decommissioningof SMRs. For SMRs to be sustainable, it is necessary to plan for decommissioning and management ofspent fuel and waste early in their development; this ensures the right infrastructure exists for spent fueland waste management, facilitates confidence of interested parties, and avoid unwanted liabilities andlegacies for future generations.
There are many different technologies being advanced and promoted throughout the world to fill the
SMR demand. The adopted approaches depend on the particular SMR design, which may give rise tospent nuclear fuel (SNF) and radioactive waste (RW) either with well-known properties or, on thecontrary, to SNF and RW for which new processing steps need to be developed. The approach to
managing these will depend in part on a country’s already existing spent fuel and radioactive waste
management plans and practices and may need to be adjusted in case additional processing steps areneeded.
One central lesson learned from prior deployment of innovative nuclear technologies remains the needto fully understand the back-end liabilities that may arise, and to have credible management paths
available – otherwise the promise of sustainable deployment cannot be upheld.
Of the 83 SMR designs reported in this booklet, many of them are still in the conceptual design phase.
Given the early stage of development, the key question is how far can SMR designs ensure that SNF
and RW arising throughout the lifecycle of the plant – construction, operation and decommissioning –can be kept to a minimum; and can be managed as much as possible with tried and proven technologies.
This is particularly relevant for designs which may generate new spent fuels and radioactive waste
streams or for which disposition paths may not be readily available.
The existing national policy and strategy of a country are also significant in determining spent fuel(SFM) and radioactive waste management (RWM) approaches. Establishing responsibilities and funding
mechanisms, as well as setting out technological and programmatic decision making on SFM and RWM
are all essential considerations for the developers of SMRs. This becomes most relevant for countries
which have no prior experience with nuclear power generation and had no need to have a strategy for
SFM and RWM.
The safe, secure and sustainable management of spent fuel and radioactive waste arising from theoperations of nuclear power reactors, as well as from associated nuclear fuel cycle activities andultimately from the decommissioning of these nuclear power reactors is key to the future of nuclear
energy. The nuclear fuel cycle approach (addressed in the previous annex) is a major factor influencing
the type of spent fuel and waste and associated properties that need to be managed; and that in addition
to the specific options for reactor design and operations and ultimate decommissioning of the SMR, also
influence what type of waste and waste forms are generated (i.e., waste is not limited to decision of
direct disposal or reprocessing).
Technical solutions already exist ranging from reprocessing and recycling, to conditioning of the typesof spent fuel currently in use, to the knowhow how to dispose spent fuel and high-level waste in deep
underground repositories. Partitioning and transmutation are advanced processes which have thepotential to further reduce the impact of nuclear waste to be disposed of in future. A number of facilities
in operation are effectively managing and disposing of low-level waste (LLW) and low- andintermediate-level short-lived waste (LILW-SL).
The emergence of SMRs presents an opportunity to demonstrate a new paradigm, in which effective
SFM and RWM should be considered and incorporated early in the conceptual design stage. Solutions
385

for managing spent fuel and radioactive waste arising from SMRs will be one of the important factors
for Member States to consider when selecting a technology. This approach will help address
uncertainties related to the back end of the fuel cycle, reduce costs and enhance public acceptance of
nuclear power. The experience built up over decades, from many types of reactors (experimental,
research and power), provides a most valuable resource for considering the SFM and RWM aspects of
SMRs. The advances in computer codes, particularly as applied in the simulation and modelling of awide range of phenomena in the design, performance and long-term conditions of new reactors systems,
is a major enabling factor in assessing SFM and RWM during the design stage. However new fuelmaterials are planned for some SMR designs which will require validation of computer codes to provetheir suitability. This requires the availability of experimental data of the new fuel materials.
Some SMR designs have features that could reduce the tasks associated with spent fuel management.
These designs adopt longer refuelling cycles of 3 to 7 years, in comparison to 1 to 2 years forconventional power plants. Some designs are even intended to operate for up to 30 years without
refuelling. However, even in such cases, there will be some spent fuel left, which will have to be properly
managed. For most of the innovative non-water cooled SMRs, the designers propose to either use theexisting infrastructure or to adapt it for the new spent fuel and radioactive waste stream. Countries withestablished nuclear power programmes have been managing their spent fuel for decades. They havegained extensive experience and have proper policies, strategies and infrastructure in place that mightenable the management of spent fuel arising from SMRs based on technologies of the current fleet. Since
specific plans and limits on a final repository may already be in place it may impose restrictions and abarrier to some of the new fuel cycle options and its specific waste forms.
Land-based water-cooled SMRs
Land-based water-cooled SMR developers in general adopt SFM and RWM plans similar to that ofoperating advanced water-cooled reactors. Facilities for waste treatment and storage are provided. For
instance, in SMART, that is planned for deployment in Saudi Arabia, the liquid radioactive waste will
be processed through demineralizer package to minimize the shipment volume of solid waste; while thegaseous radwaste system performs holdup and release in a controlled manner. In water-cooled SMRs,
operation without soluble boron in the primary coolant allows a significant reduction in theenvironmental discharges and concomitant simplification of the waste treatment systems.
Advances are also being made in the dry storage technologies. Holtec International, the developer of the
SMR-160, has developed a Multi-Purpose Canister called MPC-37 which among other inventories could
also be used to store the spent fuel coming from the SMR-160 in future. This canister is designed for
the HI-STORM UMAX modules, an underground vertical storage cask design.
For decommissioning, SMR design organizations also plan systematic dose reduction to personnel indecontamination and decommissioning. In integral PWRs, the longer distances from the fuel, and the
additional material and components between the active core and the RPV, can decrease the fast neutron
fluence on the RPV by a significant factor compared to a loop-type PWR. This essentially eliminates
vessel embrittlement, the need for surveillance coupons for periodic in-service inspection of the RPV. It
will also reduce the activation of the steel components significantly.
Marine-based water-cooled SMRs
In the case of Russian’s floating nuclear power plants (FNPP), onsite refuelling is not required when
deployed in a host state. The spent fuel is initially stored on board the FNPP and then will be taken backto the supplier for processing in the Russian Federation. In the RITM-200M, the waste is stored within
the OFPU, not in the operation site water area. The waste ensuing from plant operation is compact, hasa low activity level and is securely isolated from the biosphere. It has been verified that there is no effectto marine organisms in the deployment site water area.
In the case of ACPR50S, liquid radwaste system is designed to prevent or minimize the creation ofradioactive liquid effluent, and this achieved, wherever possible, by internal recycling. Gaseous
radwaste system is designed to minimize the radioactivity associated with the resulting environment

386

discharge. Solid radwaste system is designed to collect, preliminarily treat and temporarily store thesolid wastes during the operation. And the wastes will be sent to disposal site for final disposal.
In the case of BANDI-60, a floating power plant, the spent fuels are transported from and to an inland
fuel fabrication and storage facility using a dedicated fuel transport ship. The back-end fuel cycle option
for spent fuel is dependent on the plant owner’s policy and requirements.
HTGR-type SMRs
Due to higher thermal efficiencies and increased burnup, HTGRs produce about 40% less high-level
waste per unit of energy produced, including significantly less plutonium, especially just one quarter ofthe 239Pu content compared to a single-pass typical LWR cycle (mono-recycling). The storage anddisposal requirements largely depend on the volume, activity and fission product decay heat, that could
be up to 50 times lower (per volume) for HTGRs as compared to LWRs. In this form, proponents are
exploring options different from conventional packaging due to the lower specific source term and heatproduction generation) and with radioactivity already contained in coated particles with a stable siliconcarbide layer (that will last for more than a million years).
In comparison, the HTGR spent fuel volumes are much larger requiring much more space. If disposedof as part of a larger waste programme (designed for LWRs), the disposal of HTGR spent fuel spheres
or elements may therefore take up a relative larger part of the facility than expected. For example, thefuel volumetric content of uranium in a pebble fuel sphere is significantly less than 1%. So even thoughspent fuel spheres may provide a stable multi-barrier containment with very low heat generation, volume
reduction may be attractive. Separation of the coated particles from the graphite matrix (or fuel block)then become a more attractive option since the coated particles present excellent radioactivity
containment characteristics also for long term storage or disposal. Reprocessing is also an option.
Although many studies have also been performed to dispose of, condition or reprocess the large volumesof graphite from past gas-cooled reactor projects, none has been implemented on commercial scale anddirect disposal may be the preferred option.
In general, HTGR-type SMRs have a dedicated waste handling system at the reactor site to store lowand medium-level liquid and solid radioactive waste generated in all operating modes. For
decommissioning, spent fuel spheres are removed from the spent fuel tanks and loaded into the spent
fuel transport casks for final disposal.
The two fast spectrum high temperature gas-cooled reactors – EM2 and FMR, both have radioactive
waste management systems to collect, process and store radioactive materials from plant operating
systems, and storage facilities prior to being disposed or recycled.
Liquid metal-cooled fast neutron spectrum SMR
The fast neutron spectrum reactors are well known for their potential to substantially reduce the burdenof generated nuclear waste through the burning of Pu and Minor Actinides recycled as new fuels. Thisis still a major consideration in pursuing fast spectrum SMRs but other aspects such as very long fuelcycle is also important. The sodium-cooled 4S SMR simplified-plant design contributes to waste
reduction during operation and decommissioning. Some of the LMFR-type SMR designs also
progressively apply the radiation-equivalent approximation (in relation to natural occurring raw
materials background doses) in its radioactive waste disposal studies. This requires the development offuel that can accommodate the recycling of the minor actinides. For the design ARC-100, long term
storage is planned in containers that meet the requirements for the deep geological repository design of
the Nuclear Waste Management Organization (NWMO).
Molten salt SMRs
In all the (thermal spectrum) MSRs included in the booklet, the gaseous fission products are actively
removed and stored (to decay). Several designs propose the removal of the fuel salt (after cooling) to acentral facility for reprocessing or conditioning and disposal. So far there are no routes for spent saltclean up that requires pyro processing technologies and disposal. The residual fuel waste will be387

transported to geological repositories when the plant is in the decommissioning stage. Actinides such as
U/Pu/Th/MA separated at the off-site reprocessing facility are recycled to MSR and the FPs and salt will
be stored to cool down at the disposal facility. In this way (in theory) the lifetime of the waste can be
reduced to the few hundred years to decay (excluding the few very long-lived fission products).
Microreactors
In the case of microreactors, spent fuel and radioactive waste management approaches mostly followthat of the reactor family and coolant-type. For example, the Fully Ceramic Micro-encapsulated(FCM™) fuel from the MMR design may be treated similar to the HTGRs coated particle fuels.
Research and development on the spent fuel and radioactive waste management of some unique fuelforms should be further pursued.

388

389

Coated particle separation fromgraphite will reduce volume by
up to a factor of 100.

- Low and intermediate level
waste from plant operation will
be conditioned by different
process technologies.
- Possible graphite recycling, 14C
separation process- With higher thermal
efficiencies the radiotoxicity anddecay heat will be lessened by
50% for HTGRs as compared to
LWRs.
- Dry storage with naturalconvection after short material
active cooling.
- Facilities for long-term storageof spent fuel and solid radwaste
are in the NPP complex- Spent fuel take back is to date
not considered in HTGRs.
- With TRISO fuels, the core has
low power density. If affordable,
accident tolerant fuel can beused.

Volume reduction of all waste
types are the common principle.

For land-based water-cooled
SMRs: similar to that ofoperating advanced WCRs,
using the available technology
solution.

- Approach are very similar tothat of the current fleet of large
LWRs.
- In general, designers haveplans to safely store, handle anddispose of all the spent fuelincluding the on-site storage.

- For marine-based water-cooled
SMRs deployed in hostcountries, spent fuel will betaken back to the vendor’s
country, e.g. Russian Federation.
- For embarking countries, spent
fuel processing can be done in
the reactor supplier’s country.

Waste
Processing

Storage
Approach,
Spent Fuel
Pool Cooling
Mechanism

Spent FuelTake-back
Option,Marketpotential

VolumeReduction
andConditioning

HTGR-type SMRs

Spent fuel take back is to date
not considered in LMFR-type
SMRs.

Spent fuel will be stored until
reprocessing and fuel cycle
closure becomes economically
viable.

The required processes havebeen studied by advanced
countries as part of their (large)
fast reactor programmes.

- Simplified designs contributeto waste reduction.
- Development of fuel to include
recycled minor actinides toreduce spent fuel lifetime.

Liquid metal-cooled
fast neutron spectrum
SMRs

For some MSR designs, if
deployed in host countries, spent
fuel and reactor modules can betaken back to the vendor’s
country,- Development of special fuelcask or fuel is cooled within thereactor (tank or pot).
- KP-FHR with TRISO pebble
fuel, the waste is packaged in
multi-purpose canisters for dry
interim storage and subsequent
off-site transportation for direct
geologic disposal or recycling.

Gaseous fission products are
removed during operation andstored onsite to decay.

- Actinides such as U/Pu/Th/MA
separated at the off-site
reprocessing facility to berecycled.
- Fission products and salt are
stored to decay or disposed of.

Molten Salt SMRs

SMR designs by type of coolants and technology characteristics

Water-cooled Reactors

WasteManagement
and Disposal
Plan
Categories

Table VII-1 Waste Management and Disposal Plan adopted by SMR Designs

Most designers propose fuel tobe handled by central facility.

Typically, microreactors havelife-time core loading andtherefore no storage in the plant.

All radioactive waste generated
from operation will betransferred to the designatedwaste area, to be categorized andpackaged for removal from the
NPP site.

Small volumes relative to rest of
nuclear programme for nuclear
power nations.

Microreactors

ANNEX VIII
Enabling Design Features to Facilitate SMR Decommissioning

Introduction
Decommissioning of nuclear power facility is an activity at the last phase of the nuclear facility
lifecycle, which is unavoidable. It does not produce operational economic income and generally
requires an extended period for its completion. Therefore, successful implementation of
decommissioning is one of the major factors for realising low cost of nuclear power generation and will
depend on various considerations including the initial design. In addition, taking decommissioning into
account from the design phase will increase public acceptance of the development of new reactors.
Design and operational regimes of SMRs might benefit from accumulated decommissioning
experiences, best practices and lessons learned to optimize the decommissioning process. Specificinnovative features, such as modular design, selection of advanced low impurities construction
materials or easier decontamination access should make the decommissioning process more effective,
faster, limiting/optimizing secondary waste and less costly. However, some challenges will remain,
including development of specialized dismantling tools and remote handling equipment or multi-facility
decommissioning of SMRs at the same site. This should be considered at the design phase to have the
SMRs technology better adapted for the future decommissioning needs.
In addition to the SMRs technical aspects, the long-term liability in terms of decommissioning might
be well supported through the proper decommissioning planning and costing, record keeping of
documents important for decommissioning, digitalization for decommissioning etc. Circular economy
opportunities might be open considering possible reuse of SMR modules or their locations for new
reactors, especially if the preferrable immediate dismantling strategy will be applied. What willcontinue to be crucial in the next decade is the effective integration of decommissioning operations andmanagement of specific waste arisings from SMRs.
A large number of factors were found which should have been considered or reflected in the design
stage in decommissioning of NPPs and research reactors. In addition to that should decommissioning
experience and lessons learned from the current nuclear facilities such as nuclear power plants or
research reactors support new designs of SMRs. Some considerations implemented on the design stage
could lead to minimising an economical loss or contribute to the safe and more effective activities
during the decommissioning phase.
Decommissioning Aspects of SMR Design
The IAEA is engaged in efforts with Member States to develop an understanding of steps being takenin the design phase of SMRs to address decommissioning needs.
The launch of a Survey was decided in 2019 during the first Consultancy meeting on SMRs Design for
Decommissioning. The first draft was provided by the invited experts. In 2020, during the second
Consultancy meeting on SRMs Design for Decommissioning, the draft was updated. In June 2021, thesurvey was finalized after some tests within the Agency, with the assistance of the NPTDS (Nuclear
Power Technology Development Section). Then, 128 requests were sent to designers, vendors and otherstakeholders. 22 replies from 18 Member States were received by 3Q 2021.
The analysis done by the Agency shows that:
(i)

390

The most relevant topics (priority in choosing the first proposal answer): 68% of the repliers
will periodically update their preliminary decommissioning plans, 55% of the repliers havedecommissioning specialists in their organizations, 42% of the repliers have examined the

(ii)

(iii)

advantages and disadvantages of their coolants and moderators to limit the waste
production.
Consensual answers: the periodical updates of the preliminary decommissioning plans, theestimation of the amounts of radioactive waste generated during decommissioning, theconsideration of waste management cost in the overall cost of SMR lifecycle, theconsideration of standard tools or specifically developed techniques as part of design, theaccess to the lessons learned, the feedback of lessons learned from power reactors andresearch reactors, and the number of decommissioning specialists in the organizations.
Due to the diversity of the projects, answers without a clear trend: the design of tools or
robots for maintenance or decommissioning, the design of paths to remove reactor modules
in one piece, and the transportation of modules to external dismantling workshops.

The survey provides a positive initial overview of SMRs design for decommissioning. The analysis of
the replies shows relevant information and interesting optional comments. Furthermore, it wassuggested that the second “advanced” questionnaire will be administered in 2022 based on the
additional ideas and suggestions raised by participants of Technical Meeting in November/December
2021 at the IAEA.
Additionally, there was discussion around the possibility of considering the aspect of maintenance ofsystems needed after shutdown (ventilation, handling systems, pipes, valves, etc.), fuelling/defuelling,radiological shields to remove big components or structures (or to protect workers during maintenance)in the design stage. System and tools that will be used during decommissioning must be identified tokeep them operational. Also mentioned was the importance of the costing estimation and its updateeven after shutdown as well.
Decommissioning Lessons Learned and Aspects Related to SMR Designs
The general rule for the effective decommissioning of all nuclear facilities including SMRs is to benefit
from the “decommissioning-friendly” design. In case of SMR, the nuclear island could be considered
as exchange module which could be transported to the factory of origin. This approach could reduce
costs and scope of activities on the locality, increase nuclear and radiation safety and acceptance of the
SMR use. In general, the adopted decommissioning approaches depend on the particular SMR design
and more importantly on a country’s existing spent fuel and radioactive waste management plans andpractices.
Most SMRs are still in the conceptual design phase. The existing national policy and strategy of acountry are also significant in determining radioactive waste management (RWM) approaches.
Establishing responsibilities and funding mechanisms, as well as setting out technological andprogrammatic decision making on RWM, are all essential considerations in whole decommissioning
and back-end part of SMRs existence.
Technical and organizational aspects of SMRs specific problems might range from the facility design
details including modularity, selection of construction materials etc. to how to be better prepared interms of decommissioning planning and how to apply or adapt existing decommissioning strategies inline with the radioactive waste management. There might be also considered potential forstandardization of some decommissioning activities (e.g., full-scope decontamination or dismantling ofprimary equipment) after the operational lifetime will be over.
There were different lessons learned accumulated from the past activities of nuclear facilities
decommissioning which were incorporated, e.g., to IAEA-TECDOC-1657 “Design Lessons Drawnfrom the Decommissioning of Nuclear Facilities”. Design for decommissioning is also one of the
aspects required in many new Safety Standards.

391

General Discussion on SMRs for Decommissioning
The knowledge of the chemical compositions (including impurities) of the used material is required to
calculate the radiological inventories of activated waste. Because the cost of chemical analyses is lower
for non-activated samples, the sampling should be done prior to operating the reactor. The sampling
can also be done by the steel manufacturer to reduce the cost and at the same time to provide analyses
to measure with a high sensitivity a wide range of nuclides. As an example, Co-60 will impact the dose
rate of the workers and the shielding that will be used during the decommissioning and the waste
production. To limit Co-60 arisings, the concentration of metallurgy Cobalt should be limited.
Other ‘difficult to measure’ radionuclides could directly need to be also considered if they impact theclassification of the waste. Some of them are produced by neutron activation of impurities ignored by
the steel manufactures.
An important outcome of the experience of NPPs under decommissioning and NPPs still in operation,
but approaching the decommissioning phase, is that an accurate, detailed and complete radiologicalactivation inventory of structure and components needs to be completed. It allows an optimization of
dismantling activities considering waste management strategy and radioprotection measures. In
particular, it could overcome the recourse, after the shut-down, to perform laborious activation
characterization campaigns, avoiding workers exposure and additional cost, allowing to performconfirmation measurements directly during the on-going dismantling activities. In addition, theknowledge of structures and components activation during the operation could be useful for
maintenance program and also in case of need of some component/structure removal for replacement.
Basically, the activation of structure and components of the reactor could be obtained by the numericalsimulations that require:







3D model of the reactor, with a detailed representation of structures inside and outside thevessel, and different temperatures of the primary coolant;
3D neutron sources models;
3D multigroup neutron flux map on the structures of interest taking into account as much as
possible different neutron source distributions (different power distributions) reflecting
different loading pattern, average control rods position, average absorber concentration etc, orat least significant power distribution variations;
Calculation of the activities for each structure or component of interest, taking into account theexact chemical composition of material paying attention to impurities concentrations;
Validation and continuous improvement of the calculation scheme and model comparing resultwith measurement of detectors/samples such as for example that ones of vessel fluence
monitoring program.

It could be useful to consider these aspects from the earliest phases of design, construction and operationof SMRs (and of course in general for any other future reactors). During the design phase, it could be
useful to foresee an integration of the power and vessel fluence monitoring program implementing
suitable detectors/samples in locations of interest inside and outside the vessel properly oriented to theactivation inventory evaluation. During the construction it would be necessary to collect and record all
the information regarding chemical composition of all structures and components materials withparticular attention to impurities concentration. Also during the operation phase, it could be important
to continuously tune the calculation scheme and model considering the different operating conditions
including the shutdown phases.
Concerning the cost and the funding, the holistic view of the life cycle must be considered (design,building, operating and decommissioning). A work has been done by OECD/NEA in this regard to392

propose a methodology for a comprehensive approach to determine cost covering full life cycle,
including for SMRs.
The decommissioning strategy depends, besides other factors, on the existing / planned waste routes
and the design of nuclear facility. If the waste routes are not known or final disposal not existing, theimmediate dismantling will be practically impossible. The use of particular / specific material will
impact the waste route. The design and accessibility of components will determine thestraightforwardness of the decommissioning initiative. It is important to keep in mind that cost, strategy
and available waste routes are all interconnected factors when tackling decommissioning projects.
Concluding Remarks
According to the IAEA’s Advanced Reactors Information System (ARIS), there are over 80 SMR
designs and concepts globally at various developmental stages. Some of them are on the advanced
stages of construction and claimed as being near-term deployable. The SMR related research anddevelopment are also being conducted in the existing and newcomer nuclear energy countries.
The initiative to incorporate features that consider decommissioning needs into SMR design is a great
opportunity for the nuclear industry to further increase the appeal of these revolutionary technologies.
Addressing such issues from the conceptual design stage will allow for an easier and much more
efficient decommissioning when that time arrives.
The ‘plug and play’ nature of small modular reactors may allow for the ‘unplugging’ and storage of an
entire module upon decommissioning, thus reducing much of the energy and material demands of thedecommissioning process. Modularisation could enable opportunities for improving SMR
decommissioning. A novel topic potentially improving SMR decommissioning is the link between
modularisation and circular economy.

393

ANNEX IX
Experimental Testing for Design Verification and Validation
Introduction
SMRs from major types of technologies, including WCRs, HTGRs, LMFRs, MSRs and microreactors
may adopt innovative features that involve or introduce novel physical phenomena. Advanced reactor
designs including SMRs propose design solutions that have not been adopted before and whichsometimes are not yet supported by extensive operational experience and experimental data. A numberof designs propose design solutions that have to be supported by extensive operational experience andexperimental data. Considering those factors, new scientific modelling combined with highperformance computing must be developed to solve complex new mechanisms, phenomena, andprocesses occurring in advanced reactors. Design tools and computer codes for safety analysis require
verification and validation (V&V) for the purpose of licensing or design certification. Experimental
testing is conducted for design verification and validation, particularly for computer codes used forsafety analyses to support licensing and/or design certifications. Design verification is performed toensure the design complies with customer requirements, technical requirements, regulatory
requirements, and codes and standards.
Water-cooled SMRs
Integral PWR is a major type adopted for SMRs, where the primary system is integrated in the RPV.
Several near-term deployable SMRs are of iPWR-type. Several tests are required for validation, design
certification and deployment of SMRs. The impediments in the deployment of SMRs need to be resolved
through testing and qualification of components, R&Ds, training and international collaboration.
Integral effects test (IET), separate effects test (SET) and component effect test (CET) facilities arefundamental for the licensing purposes. The SET facilities and related testing are important todemonstrate both the safety and the performance issues of innovative components (e.g., internal
CRDMs, internal pressurizers, internal SGs. SET and IET are the major tests required for the validationof innovative technology and design, which includes – thermal-hydraulic test, computer models andcode validation used to predict the thermal efficiency, performance, and safety of a reactor etc. The
integral thermal-hydraulic test examines the system interaction in the design. While the data from theSETs are being used to develop and verify safety analysis model, the IETs is being used to verify thecapability of the analysis method and predict the integrated innovative safety systems. The SET and IET
performed for different SMR design with focus on iPWR-type is discussed below.
CAREM developed by CNEA, Argentina
A full-scale loop has been built in order to test the innovative control rod drive mechanisms. This facility
is operating at the same parameters (pressure and temperature) as RPV plant design conditions and was
designed for reactivity adjustment and control rods calibration. It will also be used to verify the
behaviour of the fast shutdown rods (FSS).
ACP100 developed by CNNC, China
Seven verification tests were completed for CRD line cold and hot test, CRD line anti-earthquake test,
Internals vibration test research, Fuel assembly critical heat flux test research, passive ECCS integration
test, CMT and passive RHR system test, and passive containment heat removal test.
SMART developed by KAERI, Republic of Korea
The advanced design features of SMART were verified by a comprehensive technology validation
program that includes safety and performance tests. The safety tests consist of core CHF tests, SETs and
IETs of the safety systems, thermal-hydraulic experiments, and digital man-machine interface system(MMIS) tests. The performance tests covered fuel assembly out-of-pile tests, performance tests of the394major components including RPV dynamic tests, RCP moc-kup test and SG irradiation test, and MMIS
control room tests.
VOYGRTM developed by NuScale Power Inc., United States of America
NuScale has designed, built, and operates a one-third scale prototype testing facility that replicates theentire NPM and its safety systems including the reactor building cooling pool. It provides an electricallyheated core to bring the system up to operating temperature and pressure. This facility is used to test the
NPM and gather data for thermal hydraulic codes, safety analysis code, and reactor design validation.
NuScale has designed and performed comprehensive testing to validate the operation of the helical coilsteam generators, the safety valves, the control rod assembly and drive shafts, fuel and various othersystems. NuScale testing programs have been audited by the NRC.
HTGR-type SMRs
HTR-10 developed by INET, Tsinghua University, China
The HTR-10 is the fundamental test reactor essential to develop the HTR-PM. From 1986 to 1990,
eight (8) research topics for key technologies were defined: (i) a conceptual design and the supporting
reactor physics and thermal fluid design and safety software codes; (ii) manufacturing process of thefuel spheres; (iii) reprocessing technologies for the thorium-uranium cycle; (iv) core internal graphite
structure design and supporting analysis; (v) helium technology establishment, (vi) pressure vessel
designs, (vii) the fuel handling design; (viii) development of special materials. The following
engineering experiments are needed prior to the commissioning: (i) a hot gas duct performance test; (ii)measurements to establish the mixing efficiency at the core bottom (limit stratification and heat streaks);(iii) two-phase flow stability tests on the once-through steam generator; (iv) fuel handling performance
test; (v) control rods drive mechanism performance; (vi) V&V of the digital reactor protection systems;(vii) measurements to confirm the neutron absorption cross-section of the reflector graphite and (viii) a
performance test for the helium circulator.
GTHTR300 developed by JAEA, JapanThe test results using the HTTR will be utilized for realization of GTHTR300. The test items cover fuelperformance and radionuclide transport, core physics, reactor thermal hydraulics and plant dynamics,
and reactor operations, maintenance, control, etc. The GTHTR300, which is meant for commercial unit,
shall demonstrate its ability to operate in normal cogeneration mode or with electric or hydrogen system
operating alone in such a case as forced shutdown of either system. The results of the system
performance analysis showed that the reactor could be continuously operated with the above variableload conditions. However, an actual demonstration test is warranted for performance confirmation.
Liquid Metal-cooled Fast Neutron Spectrum SMRs
BREST-OD-300 developed by NIKIET, Russian Federation
The complete detailed design of the BREST-OD-300 reactor facility has been carried out. To date,experimental justification of components, elements and equipment of reactor facilities has been carriedout using small- and medium-scale mock-ups and pilot models. Verified and certified software tools
were used for computational design justification.
SEALER-55 developed by LeadCold, Sweden
An electrically heated prototype of SEALER-55 will be built and operated in Oskarshamn, Sweden forthe purpose of validating its safety concept, operational and maintenance procedures, as well as materials
performance. The prototype is designed with a power of 3 MW, produced in 7 heated rod assemblies
with 37 rods each. The height of the prototype is 1:1 with respect to the SEALER-55 to validate residual395

heat removal capability by dip-coolers, whereas the diameter of the vessel is scaled by 1:2.5. The
prototype is under engineering design, with the intent to have it in operation by 2024.
Molten Salt SMRs
IMSR400 developed by Terrestrial Energy Inc., Canada
TEI has completed the first phase of testing to confirm Fuel Salt thermo-physical properties. The tests
accounted for Fuel Salt aging - build-up of fission products. Verification of the experimental findings’
reproducibility will follow. Testing of Fuel Slat redox potential and respective interfacing material
response is in-progress. Solubility limits of trifluorides and iodides inside the salt are in progress. Salt
chemical and mechanical interactions with non-irradiated graphite is completed. Analogous test using
the irradiated graphite specimens will follow. Irradiation of selected graphite grades is in progress andwill conclude with the property-testing of the irradiated graphite. The test devoted to the irradiated alloy
property testing entered its detailed design phase. Alloy corrosion testing is in progress. Waste
management feasibility study is completed, and experimental phase is being planned. Reactor physics,
Thermal-hydraulics and instrumentation tests are being designed and planned.
Microreactors
MARVEL developed by Idaho National Laboratory, United States of America
The MARVEL team conducts rapid prototyping tests to mature its technologies. So far, more than ten
separate effects tests have been conducted, including, but not limited to, the intermediate heatexchangers, control drums, instrumentation control, neutron detection, shutdown rod actuators, andStirling engines. Due to MARVEL’s novel thermal hydraulic circuit, where a liquid metal natural
circulation primary loop is in series with four parallel liquid metal natural circulation loops, an IET ofthe system was considered necessary for verifying the transient dynamics of the system before reactor
construction. Hence the team has successfully designed and fabricated a full, scale electrically heatedprototype of the MARVEL reactor. The test hardware includes (i) a Full-scale mechanical IET test
article; (ii) eight electrical control cabinets; (iii) a structural frame; (iv) four IET flow meters; (v) morethan 200 thermocouples and pressure transducers; and (vi) four Stirling engines, engine control, andheat rejection units. The goals of this IET are to: (i) validate flow and heat transfer characteristics of
MARVEL technology; (ii) benchmark modelling and simulation parameters; (iii) streamline
manufacturing methods; (iv) de-risk supply chain; and (v) train operators.
HOLOS-QUAD developed by Holos Gen, United States of America
The design has been scrutinized for approximately 4 years by design and resource teams formed byscientists and subject matter experts at the national laboratories, industry and academic institutions,
sponsored in part by the U.S. DOE ARPA-E MEITNER funding program. Validation activities under
this program included technical and economic feasibility executed with high-fidelity codes developedby the national laboratories. The results of these activities were published through numerous peerreviewed technical papers addressing, thermal-hydraulic, heat transfer, load-following, transient
analyses, loss of coolant, loss of electrical power, flooding, passive decay heat removal and structural
aspects of the design. As part of validation activities, costing analyses of the design's SSCs' life cycle
were executed including upfront cost for decommissioning, spent fuel transport and storage for 50 years.

396

397

~30 MW(e) / 100
MW(t)

CAREM

Notes:The design
description of IRIS
was reported in theprevious editions of
ARIS Booklet from2012 to 2020.

335 MW(e) / 1000MW(t)International
Reactor Innovativeand Secure (IRIS)SMR Designs,
Power Rating

The integral reactor coolant
system consists of 8 helicalcoil SGs, internal CRDM andintegral pressurizer installed
within the RPV.
Heat removal from inside thevessel depressurizes theprimary system bycondensation; effective heatremoval by SG’s emergency
heat removal system (EHRS).

SGs safety valves eliminated
Passive EHRSPassive containment cooling.
Auxiliary building is fully
seismic isolated.
Integral PWR – primary coolant
system within the RPV, selfpressurized, in-vessel hydraulic
CRDMs, primary coolant flow
relies entirely on natural convection
and safety systems relying on
passive features.

Specific Features
Full scale test of steam
generator helical-coil tubes.
EHRS passive safetysystems test scaled on
power/volume.

Fuel element, already tested in
the RA-6 reactor.

Component Effect Tests

High-pressure and high-temperature
rig test for the innovative hydraulic
CRDMs.
Thermal limits and CHF tests.
Low pressure loop: hydraulic losses
and flow vibration test.
CHF facility at Pica, Argentina forneutronic code validation.










The containment compartments are
simulated in SPES3 by separate
tanks connected, representing thedry-well, two pressure suppression
systems, two long-term gravity
makeup systems, the reactor cavity,and ADS quench tank.
The three ADS trains are simulated
in SPES3 by two trains. Each train
consists of a safety valve, a line tothe quench tank and a line to the dry
well.



Separate Effect Tests

Integral Effect Tests

Thermal-hydraulic dynamics test in conditions
similar to CAREM25 operational states (1:1 inheight and pressure, temperature – 1:335 inpower) – natural circulation and selfpressurization test.

IRIS integral test facility:
Full scale in height – temperature –pressure
Scaled 1:100 in power/volumes

Testing of most accident scenarios.
Validation of codes and behaviour ofpassive safety systems and containmentvessel coupling.
>700 measurement points, new
instrumentation developed.

Table IX-1 Examples of Tests and Tests for integral-PWR Design Developments and Licensing

398



125 MW(e) / 385
MW(t)











ACP100SMR Designs,
Power Rating

Integrated reactor design
technology
Passive severe accident
prevention and mitigation
action, such as forcontainment hydrogen
eliminator, cavity flooding etc.
A multipurpose designed forelectricity production, heating,steam production or seawater
desalination
Vertically mounted RCP
CRDM of magnetic force type
Reactor building and spent
fuel pool are located below
ground level
Passive decay heat removal
system.

Specific Features
Nuclear fuel fabrication plant at
Yibin manufactured a shortened
CF2 fuel assemblies and control
rod samples for R&D project at
Nuclear Power Institute of China(NPIC).

Component Effect Tests
Fuel assembly CHF test:In typical unit at uniform heat fluxIn typical unit at non-uniform heatfluxIn non-typical unit at non-uniform
heat fluxControl rod drive line (CRDL) antiseismic test.
Passive safety systems tests:
Passive ECCS integration test
CMT and PRHRS testFlow-induced vibration test

Test indicates that the maximum
stress was more less than high cycle
fatigue allowable stress under 100%
rated flow condition

CRDL cold and hot test

Separate Effect Tests

Core cooling and residual heat removal
system test facility (CREST) (1:1 in heightand pressure, temperature; 1:37 in volume;
1:100 in power).
CNNC/NPIC has built up comprehensive
testing facilities which fulfil the needs ofdesign, and results of the test research on
the crucial technology provided necessary
basis for finial design and safety
evaluation of the reactor.

Integral Effect Tests

399



107 MW(e) / 365
MW(t)






SMART

SMR Designs,
Power Rating

RPV contains all of the
primary components such ascore structures and SGs.
A passive decay heat removal
system in the secondary side
Horizontally mounted RCPs.
Intended for sea water
desalination and electricity
supply in newcomer countries
with small grid.

Specific Features
Basic fundamental thermalhydraulic experiments include:Boiling heat transfer
characteristics in thehelically coiled SGs.

Containment cooling systemusing heat pipe.
Verification andperformance test of
hydraulic valve
characteristics measurementtest.
The performance tests for
key parts of main coolant
pump
Irradiation test for SG tube
material (A690)

Component Effect Tests
Fuel Performance Tests:Fuel component test.
Fuel assembly out-of-pile test.

Freon CHF test.
Water critical heat flux
(CHF)/mixing test.
Safety validation tests:

Safety injection bypass test (1:5
scale).
SG and passive residual heatremoval system HX heat transfer
test (scaling ratio height/volume–
1:2.8, 1/473)
Thermal-hydraulic performance tests:
RPV assembly flow distribution test
(1:5 scale)Flow mixing header assembly
performance test (1:5 scale)Internal pressurizer/levelmeasurement test (1:6 scale)
Tests for various I&C systems
Software validation of key design
tools and methods: core physics,
core thermal-hydraulics, safety
analysis etc.

Test of flow instability in steam
generator
Test of self-pressurizer performanceTwo-phase critical flow test with
on-condensable gases

Separate Effect Tests









Thermal-hydraulic integral effect test was
carried out at VISTA-ITL test facility.
Experiment for natural circulation in
integral reactor
The integral thermal hydraulic test which
examines the system interaction in thedesign is performed.
The high temperature and high pressure
integral thermal hydraulic test facility
established performed various
comprehensive thermo-hydraulic tests.

Integral Effect Tests

400

77 MW(e) / 250
MW(t)VOYGRTMSMR Designs,
Power Rating







Natural circulation cooled;decay heat removal using
containment; built below
ground.
RPV is placed within acontainment vessel made ofhigh-strength stainless steel,submerged underwater in a
below-grade pool shared by all
modules.
100% turbine bypass
capability; dry cooling option;enhanced features for loadfollowing; island mode
capability during loss of offsite power; flexibilities forprocess heat applications.

Specific Features
Prototypic full-scale thermalhydraulic test conducted at Stern
Laboratory in Canada to obtaindata for validating sub-channel
code for design certification with
the NRC:CHF data for correlation
development.

Fuel bundle sub-channel
exit temperatures todetermine mixing
coefficients.
Fuel bundle single and twophase pressure drops.
Full-length NuScale fueldesign.

5 × 5 electrically heatedarray.

Uniform and symmetriccosine power peaking test
sections.

Demonstration of a proofof-concept inspection
system test for the helicalcoil SG using available
supplier tooling successfullycompleted.

Component Effect Tests

A full-scale upper module mock-up of thereactor is built. It includes top of thereactor module down to the elevation ofreactor vessel head. All major
components are mocked up including:The upper portion of thecontainment vessel
Major piping such as steam, feedwater, and chemical/volume control

CRDMs

Major valves such as isolation
valves and emergency core coolingvalves
Module service platform.

To obtain the necessary test data,
NuScale embarked on two major test
programs utilizing specialized facilities
at Società Informazioni Esperienze
Termoidrauliche (SIET) in Piacenza,
Italy.
A full-length SG tubes test toinvestigate important physicalphenomena and processes that
occur in the SG tubes and on the
primary side tube bank at highpressure and temperature withfocused on the secondary side
performanceThe second test focuses on overall
primary and secondary side
performance and consists of a
prototypic tube bank operated at
prototypic primary and secondary
flow conditions.

Separate Effect Tests
The NuScale integral system test (NIST-1)
facility located in Corvallis, Oregon is a 1/3
scale in volume/height and 1:1 in pressure andtemperature prototype that replicates the power
module and reactor building cooling pool. It
provides an electrically heated core to bring thesystem up to operating temperature and pressure.
Stability testing ensures that throughout
the operating conditions, stable natural
circulation is assured.

Large-scale real-time integral effects data
for LOCAs, long-term core cooling, andnon-LOCA transients to validate design
and analysis tools for design certification.

A 12–module control room simulator to
supports human factor engineering studies,
control room staffing exemption, control
room design and plant performancestudies.

Integral Effect Tests

ANNEX X
Bibliography
 Contained status, design description and mainfeatures of 32 selected SMR designs;
 Sorted by type/coolant: iPWR, PHWR, GCR, andLMFR; Sorted by Country of Origin; Included: CAREM (Argentina), FBNR (Brazil),
CNP-300 (China), Flexblue (France), IMR (Japan),
SMART (Republic of Korea), ABV-6M (RussianFederation), SHELF (Russian Federation), RITM200 (Russian Federation), VK-300 (RussianFederation), VBER-300 (Russian Federation),
WWER-300 (Russian Federation), KLT-40S(Russian Federation), UNITHERM (RussianFederation), IRIS (International Consortium),mPower (USA), NuScale (USA), Westinghouse SMR(USA), EC6 (Canada), PHWR-220 (India),
AHWR300-LEU (India), HTR-PM (China), PBMR
(South Africa), GT-MHR (USA), EM2 (USA), CEFR
(China), 4S (Japan), PFBR-500 (India), BREST-OD300 (Russian Federation), SVBR-100 (RussianFederation), PRISM (USA), G4M (USA).
 Published September 2012

 Contained status, design description and mainfeatures of 22 selected fast reactor designs;
 Sorted by type/coolant: SFR, GFR, and HLMC,
MSFR;
 Sorted by Country of Origin; Included: CFR-600 (China), ASTRID (France), FBR1&2 (India), 4S (Japan), JSFR (Japan), PGSFR(Republic of Korea), BN-1200 (Russian Federation),
MBIR (Russian Federation), PRISM (USA), TWR-P(USA), MYRRHA (Belgium), CLEAR-I (China),
ALFRED (Europe/Italy), ELFR (Europe/Italy),
PEACER (Republic of Korea), BREST-OD-300(Russian
Federation),
SVBR-100(Russian
Federation), ELECTRA (Sweden), G4M (USA),
ALLEGRO (Europe), EM2 (USA), MSFR (France).
 Published October 2013401 Contained status, design description and mainfeatures of 31 selected SMR designs;
 Sorted by type/coolant: iPWR, AHWR and HTGR; Sorted by Country of Origin; Included: CAREM (Argentina), ACP-100 (China),
Flexblue (France), IMR (Japan), SMART (Republicof Korea), ABV-6M (Russian Federation), SHELF(Russian
Federation),
RITM-200(Russian
Federation), VK-300 (Russian Federation), VBER300 (Russian Federation), KLT-40S (RussianFederation), UNITHERM (Russian Federation), IRIS
(International Consortium), mPower (USA), NuScale(USA), Westinghouse SMR (USA),SMR160 (USA),
AHWR300-LEU (India), HTR-PM (China), PBMR
(SouthAfrica),
GT-MHR(Russian
Federation),VVER-300(Russian
Federation),
RUTA-70(Russian Federation), ELENA (RussianFederation), DMS (Japan), HTR-PM (China),
GTHTR300 (Japan), MHR-T (Russian Federation),
MHR-100 (Russian Federation), PBMR-400 (South
Africa), HTMR-100 (South Africa), SC-HTGR(USA), Xe-100 (USA) Published September 2014

 Contained overview of status and main features of 18
selected large water-cooled reactor designs;
 Sorted by Country of Origin/Vendor; Included: ACPR-1000 (China), CAP-1400 (China),
CPR-1000 (China), HPR1000 (China), APR1400(Republic of Korea), APWR (Japan), AP1000(Japan),
ABWR(Japan), VVER1000(Russian Federation),,VVER1200(RussianFederation),.VVER1500(RussianFederation),
IPHWR(India), EPR(France), KERENA(France),
ATMEA1(France) , EC6 (Canada), ABWR(USA),
ESWR(USA) Published September 2015402

 Contained status, design description and mainfeatures of 48 selected SMR designs;
 Sorted by Land based and Marine based LWRs,
HTGR, Fast spectrum SMRs and Molten Salt SMRs;
 Sorted by Country of Origin; Included: CAREM (Argentina), ACP100, CAP150,
CAP200 (China), AHWR-300 (India), IRIS
(International Consortium), DMS and IMR (Japan),
SMART (Republic of Korea), UNITHERM, KARAT45, KARAT-100, ELENA, RUTA-70 (RussianFederation), NuScale, mPower, Westinghouse SMR,
SMR-160 (United States of America)
ACPR50S (China), Flexblue (France), KLT-40S,
RITM-200, VBER-300, ABV-6E, SHELF (RussianFederation),
HTR-PM (China), GTHTR300 (Japan), GT-MHR,
MHR-T, MHR-100 (Russian Federation), PBMR400, HTMR-100 SMR (South Africa), SC-HTGR,
Xe-100 (United States of America)
LEADIR-PS (Canada), 4S (Japan), BREST-OD-300,
SVBR-100 (Russian Federation), G4M, EM2 (UnitedStates of America)
Integral Molten Salt Reactor (Canada), MSTW
(Denmark), ThorCon (International Consortium),
FUJI (Japan), Stable Salt Reactor (United Kingdom),
SmAHTR, Liquid Fluoride Thorium Reactor, Mk1
PB-FHR (United States of America) Published August 2016

 Contained status, design description and mainfeatures of 56 selected SMR designs;
 Sorted by Land based and Marine based LWRs,
HTGR, Fast spectrum SMRs and Molten Salt SMRs;
 Sorted by Country of Origin; Included: CAREM (Argentina), ACP100, CAP150,
CAP200 (China), AHWR-300 (India), IRIS
(International Consortium), DMS and IMR (Japan),
SMART (Republic of Korea), UNITHERM, KARAT45, KARAT-100, ELENA, RUTA-70 (RussianFederation), NuScale, mPower, Westinghouse SMR,
SMR-160 (United States of America), ACPR50S(China), Flexblue (France), KLT-40S, RITM-200,
VBER-300, ABV-6E, SHELF (Russian Federation),
HTR-PM (China), GTHTR300 (Japan), GT-MHR,
MHR-T, MHR-100 (Russian Federation), PBMR400, HTMR-100 SMR (South Africa), SC-HTGR,
Xe-100 (United States of America)
LEADIR-PS (Canada), 4S (Japan), BREST-OD-300,
SVBR-100 (Russian Federation), G4M, EM2 (UnitedStates of America)
Integral Molten Salt Reactor (Canada), MSTW
(Denmark), ThorCon (International Consortium),
FUJI (Japan), Stable Salt Reactor (United Kingdom),
SmAHTR, Liquid Fluoride Thorium Reactor, Mk1
PB-FHR (United States of America) Published September 2018

403

 Contained status, design description and mainfeatures of 72 selected SMR designs;
 Sorted by Land-based and Marine-based WCRs,
HTGRs, Fast neutron spectrum SMRs, Molten Salt
SMRs and Microreactors;
 Sorted by Country of Origin; Included: CAREM (Argentina), ACP100 (China),
CANDU SMR (Canada), CAP200, DHR400,
HAPPY200 (China), TEPLATORTM (Czech
Republic), NUWARD (France), IRIS (IRIS
Consortium), DMS and IMR (Japan), SMART(Republic of Korea and Saudi Arabia), UNITHERM,
VK-300, KARAT-45, KARAT-100, RUTA-70,
ELENA (Russian Federation), UK SMR (UK),
NuScale (USA), BWRX-300 (USA and Japan),
SMR-160, Westinghouse SMR, mPower (USA)ACPR50S (China), KLT-40S, RITM-200M, VBER300, ABV-6E, SHELF (Russian Federation)
HTR-PM (China), StarCore (Canada/UK/USA),
GTHTR300 (Japan), GT-MHR, MHR-T, MHR-100(Russian Federation), PBMR-400, AHTR-100,
HTMR100 (South Africa), SC-HTGR, Xe-100(United States of America), HTR-10 (China), HTTR
(Japan), RDE (Indonesia)
4S (Japan), BREST-OD-300, SVBR (RussianFederation), microURANUS (ROK), LFR-AS-200,
LFR-TL-X (Luxembourg), SEALER (Sweden), EM2,
Westinghouse LFR, SUPERSTAR (USA)IMSR (Canada), smTMSR-400 (China), CA Waste
Burner 0.2.5 (Denmark), ThorCon (International
Consortium),
FUJI(Japan),
SSR-W300(UK/Canada), LFTR, KP-FHR, Mk1 PB-FHR(USA), MCSFR (USA and Canada)
Energy Well (Czech Republic), MoveluX (Japan), UBattery (UK), Aurora, Westinghouse eVinci, MMR(USA) Published September 2020

404

ANNEX XI
Acronyms
ABWR
AC
ACC
ACME
ADS
ADV
AFA
AGR
ALARA
ALARP
ARIS
ASEC
ASIV
ASME
AST
ATF
ATR
ATS
ATWS
AUSC
BCR
BDBA
BOL
BOP
BPV
BWR
CANDU
CBR
CCF
CCS
CCWS
CDF
CED
CEDM
CES
CHE
CHR
CI
CIS
CMT
CNPP
CNPP
CNSC
CNV
CPRSS
CPS
CRA
CRDM
CS
CSG

Advanced Boiling Water Reactor
Alternating Current
Air-cooled Condenser
Advanced Core-cooling Mechanism Experiment
Automatic Depressurization System
Atmospheric Dump Valve
Alumina Forming Austenite
Advanced Gas-cooled Reactor
As Low As Reasonably Achievable
As Low As Reasonably Practicable
Advanced Reactor Information System
Air Heat Sink for Emergency Cooldown
Amphora-Shaped Inner Vessel (in LFR design)
American Society of Mechanical Engineers
Auxiliary Standby Transformer
Accident Tolerant Fuel
Advanced Test Reactor
Automation System
Anticipated Transient Without Scram
Advanced Ultra-Supercritial
Back-up Control RoomBeyond Design Basis Accident
Beginning of Life
Balance of Plant
ASME Boiler and Pressure Vessel code
Boiling Water Reactor
Canada Deuterium Uranium (Canadian Pressurized Heavy-water Reactor)
Core Breeding Ratio
Common Cause Failure
Containment Cooling System
Component Cooling Water System
Core Damage Frequency
Contract Effective Date
Control Element Drive Mechanism
Containment Enclosure Structure
Containment Hydrogen Control and Filtration Exhaust System
Containment Heat Removal System
Conventional Island (Turbine-Generator Building)Containment Isolation System
Core Makeup Tank
Cogeneration Nuclear Power Plant
Cogeneration Nuclear Power Plant
Canadian Nuclear Safety Commission
Cylindrical Containment VesselContainment Pressure and Radioactive Suppression System
Control and Protection System
Control Rod Assemblies
Control Rod Drive Mechanism
Containment Structure
Compact Steam Generator

405

CSMR
CSS
CSS
CST
CSTS
CTAH
CTS
CV
CVCS
CWS
D/G
DAS
DBA
DBC
DBE
DC
DCA
DCIS
DCS
DEC
DGR
DHRS
DID
DLOFC
DRACS
DU
DVI
EAB
EBI
EBT
ECCS
ECDS
ECT
EDG
EFPD
EHRS
ENTSO
EOC
EOL
EPZ
ESBWR
ESF
ESWS
ETS
FA
FBR
FCD
FDT
FE
FEED
FGCS
FHR
FHS
FMCRD
FNPP

406

CANDU SMR
Control and Support Safety System
Control Safety System
Coolant Storage Tank
Condensate Storage and Transfer System
Coiled Tube Air Heaters
Chemical Technological Sector
Containment Vessel
Chemical and Volume Control System
Chilled Water System
Diesel Generator
Diverse Actuation System
Design Basis Accident
Design Basis Condition
Design Basis Earthquake
Direct Current
Design Certification Application (a licensing term in the United States)
Distributive Control and Information System
Distributed Control System
Design Extension Condition
Deep Geological Repository
Decay Heat Removal System
Defence in Depth
Depressurized Loss of Forced Cooling (in HTGR)
Direct Reactor Auxiliary Cooling System
Depleted Uranium
Direct Vessel Injection
Exclusion Area Boundary
Emergency Boron Injection
Emergency Boration Tank
Emergency Core Cooling System
Emergency Cooling Down System
Emergency Cooldown Tank
Emergency Diesel Generator
Effective Full Power Day
Emergency Heat Removal System
European Network of Transmission System Operators for Electricity
End-of-cycle
End of Life
Emergency Planning Zone
Economic Simplified Boiling Water Reactor
Engineered Safety Feature
Essential Service Water System
Energy Transport System
Fuel Assembly
Fast Breeder Reactor
First Concrete Date
Fuel salt Drain Tank
Fuel Element
Front-End Engineering Design
Fission Gas Collection System
Fluoride-salt cooled High Temperature Reactor
Fuel Handling System
Fine Motion Control Rod Drive (in BWR)
Floating Nuclear Power Plant

FOAKFirst of a Kind
FPC
Fuel Pool Cooling and Cleanup System
Field Programmable Gate Arrays
FPGA
FPHE
Formed Plate Heat Exchanger
Floating Power UnitFPU
FPVS
Fission Product Vent System
FRPS
First Reactor Protection System
Final Safety Analysis Report
FSAR
FSF
Fundamental Safety Function
Free Surface Separation
FSS
FSSFirst Shutdown System
FSST
Fuel Salt Storage Tank
GCB
Generator Circuit Breaker
GDA
Generic Design Assessment
Gravity Driven Cooling System
GDCS
GDWP
Gravity Driven Water Pool
Gas-cooled Fast Reactor
GFR
GHT
Gas Holding Tanks
GTG
Gas Turbine Generator
GV
Guard Vessel
HALEU High-Assay Low Enriched Uranium
HE
Heat Exchanger
HEU
High Enriched Uranium
HFE
Human Factors Engineering
HFIR
High Flux Isotope Reactor
HGDPV Hot Gas Duct Pressure Vessel
HHTS
Hybrid Heat Transport System
HIPS
Highly Integrated Protection System
HLMC
Heavy Liquid Metal-Cooled
HMIS
Human-Machine Interface system
HPB
Helium Pressure Boundary
HPCF
High Pressure Core Flooder (in BWR)HRSGHeat Recovery Steam Generator System
HTF
Heat Transfer Fluid
HTGRHigh Temperature Gas-cooled Reactor
HTR
High Temperature Reactor
HTS
Heat Transport System (in CANDU and PHWR)
HVDC
High Voltage Direct Current
HWR
Heavy Water Reactor
HX
Heat Exchanger
I&C
Instrumentation and Control
IC
Isolation Condenser
ICS
Isolation Condenser System
IHP
Integrated Head Package
IHX
Intermediate Heat Exchanger
IPIT
Intermediate Pressure Injection Tanks
IRACSIntermediate Reactor Auxiliary Cooling System
IRWST
In-Containment Reactor Water Storage Tank
ISTIntegrated System Test
IVR
In-Vessel Retention
LBB
Leak Before Break
LBE
Lead-Bismuth Eutectic coolant
LBLOCA Large Break Loss of Coolant Accident
LC
Lead Coolant

407

LCA
LCOE
LEU
LFR
LFTR
LLSF
LOCA
LOFC
LOHS
LOOP
LPFL
LWR
MA
MCP
MCR
MCSFR
MHT
MMIS
MOX
MSA
MSFR
MSK
MSR
MSR
MSRE
MSSV
MW(e)
MW(t)
MWS
MWS
NACC
NDHP
NFC
NGCC
NHP
NHSS
NI
NIS
NPP
NPS
NSSS
NTEP
NWMO
O&M
OBE
OCP
OFPU
OPEX
OTSG
OTTO
PAFS
PAR
PAS
PBMM
PC

408

Lower Containment Area
Levelized Cost of Electricity
Low Enriched Uranium
Lead-cooled Fast Reactor
Liquid-Fluoride Thorium Reactor
Low Level Safety Functions
Loss of Coolant Accident
Loss of Forced Cooling
Loss of heat sink
Loss of Offsite Power
Low Pressure Core Flooder (in BWR)Light Water Reactor
Minor Actinides
Main Coolant Pump
Main Control Room
Molten Chloride Salt Fast Reactor
Main Heat Transport
Man-machine Interface System
Mixed Uranium-Plutonium Oxide fuelMoisture Separator Reheater
Molten Salt Fast Reactor
Medvedev-Sponheuer-Karnik scale (a macroseismic intensity scale used in Russia)
Moisture Separator and Reheater
Molten Salt Reactor
Molten Salt Reactor Experiment
Main Steam Safety Valve
Mega Watt electric
Mega Watt thermal
Makeup Water System
Metal and Water Shielding
Nuclear air-Brayton Combined Cycle
Nuclear District Heating Plant
Nuclear Fuel Cycle
Natural Gas Combined Cycle
Nuclear Heating Plant
Nuclear Heat Steam Supply System
Nuclear Island
Nuclear Instrumentation Systems
Nuclear Power Plant
Nuclear Power System
Nuclear Steam Supply System
Nuclear Thermoelectric Plant
Nuclear Waste Management Organization
Operation and MaintenanceOperating Basis Earthquake
Outside Containment Pool
Optimized Floating Power Unit
Operating Expenses
Once-Through Steam Generators
Once Through Then Out
Passive Auxiliary Feedwater System
Passive Autocatalytic Re-Combiners
Passive Containment Air Cooling System
Pebbled Bed Micro Model
Primary Containment

PCCS
PCHR
PCS
PCT
PCU
PCV
PDHRS
PFB
PGA
PHTS
PHWR
PHX
PLC
PLOFC
PLS
PMG
PMS
PORV
PRA
PRHRS
PSAR
PSGC
PSIS
PSWS
PWR
QA
RAB
RCAB
RCCS
RCH
RCI
RCIC
RCP
RCPB
RCS
RCSS
RDP
RFA
RGPu
RHRS
RIA
RP
RPS
RPV
RSS
RTNSS
RV
RVACS
RW
RWST
SAS
SBO
SCADA

Passive Containment Cooling System
Passive Containment Heat Removal
System Primary Containment System
Peak Cladding TemperaturePower Conversion Unit
Primary Containment Vessel
Passive Decay Heat Removal System
Passive Feed and Bleed System
Peak Ground Acceleration
Primary Heat Transport System
Pressurized Heavy Water Reactor
Primary Heat Exchanger
Programmable Logic Controller
Pressurized Loss of Forced Cooling (in HTGR)
Plant Control System
Plant Main Generator
Protection and safety Monitoring System
Power-Operated Relieve Valve
Probabilistic Risk Assessment
Passive Residual Heat Removal System
Preliminary Safety Analysis Report
Passive Steam Generator Cooler
Passive Safety Injection System
Plant Service Water System
Pressurized Water Reactor
Quality Assurance
Reactor Auxiliary Building
Reactor Containment and Auxiliary Building
Reactor Cavity Cooling System
Reactor Closure Header
Reactor Coolant Inventory and Purification System
Reactor Core Isolation Cooling (in BWR)
Reactor Coolant Pump
Reactor Coolant Pressure Boundary
Reactor Coolant System
Reactivity Control and Shutdown System
Reactor automatic Depressurization System
Robust Fuel Assembly
Reactor-grade Plutonium
Residual Heat Removal System
Reactivity Insertion Accident
Reactor Plant
Reactor Protection System
Reactor Pressure Vessel
Reserve Shutdown System
Regulatory Treatment of Non-Safety System
Reactor Vessel
Reactor Vessel Auxiliary Cooling System
Radioactive Waste
Reactor Water Storage Tank
Small Absorber Sphere
Station Black-Out
Supervisory Control and Data Acquisition

409

sCO2
SDC
SDS
SEU
SFP
SFR
SFSC
SFSF
SG
SGL
SGPV
SIS
SIT
SLCS
SMR
SNF
SRPS
SSC
SSE
SSG
SSLC
STSG
T/G
TC
TCU
TEG
TES
TEU
TGP
TM
TRISO
TRL
TRU
UAT
UCA
UCO
UHS
UPS
URWT
V&V
VCS
VDR
WATSS
WDS
WPu
WWER
YSZ

410

supercritical CO2
Shutdown cooling
Shut Down System (in CANDU and PHWR)
Slightly Enriched Uranium
Spent Fuel Pool
Sodium-cooled Fast Reactor
Spent Fuel Storage Canister
Spent Fuel Storage Facility
Steam Generator
Steam Generator in Liquid Region
Steam Generator Pressure Vessel
Safety Injection System
Safety Injection Tanks
Standby Liquid Control System
Small Modular Reactor
Spent Nuclear Fuel
Second Reactor Protection System
Systems, Structures and Components
Safe Shutdown Earthquake
IAEA Specific Safety Guide
Safety System Logic and Control
Spiral-tube steam generators (in LFR design)
Turbine/Generator
Turbo Compressor
Thermal Conversion Unit
Thermoelectric Generator
Thermal Energy Storage
Thermoelectric Unit
Turbine Generator Package
Turbo MachineTri-structural Isotropic particle fuelTechnology Readiness Level
Transuranic waste
Unit Auxiliary Transformer
Upper Containment Area
Uranium Oxy Carbide
Ultimate Heat Sink
Uninterrupted Power Supply (System)
Ultimate Heat Sink and Refueling Water Tank
Verification and Validation
Vessel Cooling System
Vendor Design Review (a licensing term in Canada)
Waste to Stable Salt
Waste Disposal System
Weapon-Grade Plutonium
Water-cooled Water-Moderated Power Reactor (Russian PWR)
Yttria Stabilized Zirconia pellets

Annex XII
Acknowledgement
Organisations in Member States
Comisión Nacional de Energía Atómica (CNEA), Argentina
Agencia Boliviana de Energía Nuclear (ABEN), Bolivia
ARC Clean Energy, Canada
Candu Energy Inc. (member of the SNC-Lavalin Group), Canada
Terrestrial Energy, Inc., Canada
Moltex Energy, Canada
China National Nuclear Corporation (CNNC), China
China General Nuclear Power Corporation (CGNPC), China
Shanghai Institute of Applied Physics (SINAP), Chinese Academy of Sciences (CAS), China
Shanghai Nuclear Engineering Research & Design Institute (SNERDI), State Power Investment
Corporation (SPIC), China
Tsinghua University, China
China Nuclear Energy Association, China
Centrum výzkumu Řež, Czech Republic
UWB Pilsen & CIIRC CTU, Czech Republic
Seaborg Technologies ApS, Denmark
Copenhagen Atomics, Denmark
Électricité de France (EDF), France
Jimmy Energy SAS, France
National Research and Innovation Agency (BRIN), Indonesia
newcleo srl, Italy
Politecnico di Milano, Italy
Hitachi-GE Nuclear Energy, Japan
Japan Atomic Energy Agency (JAEA), Japan
Mitsubishi Heavy Industries Ltd., Japan411

Toshiba Energy Systems & Solutions Corporation, Japan
International Thorium Molten-Salt Forum (ITMSF), Japan
THORIZON, Netherlands
KEPCO E&C, Republic of Korea
Korea Atomic Energy Research Institute (KAERI), Republic of Korea
Korea Hydro & Nuclear Power Co., Ltd. (KHNP), Republic of Korea
Ulsan National Institute of Science & Technology (UNIST), Republic of Korea
JSC “Afrikantov OKBM”, Rosatom, Russian Federation
NIKIET, Russian Federation
National Research Centre ‘Kurchatov Institute’, Russian Federation
JSC AKME Engineering, Russian Federation
King Abdullah City for Atomic and Renewable Energy (K.A.CARE), Saudi Arabia
ESKOM Holdings SOC Ltd., South Africa
PBMR SOC Ltd., South Africa
STL Nuclear (Pty) Ltd., South Africa
LeadCold, Sweden
STAR ENERGY SA, Switzerland
Moltex Energy, United Kingdom
newcleo, Ltd, United Kingdom
Rolls-Royce SMR Ltd., United Kingdom
Urenco, United Kingdom
Flibe Energy, Inc., United States of America
Framatome Inc., United States of America
GE-Hitachi Nuclear Energy, United States of America
Kairos Power LLC, United States of America
Holtec International, United States of America
Idaho National Laboratories (INL), United States of America
NuScale Power Corporation, United States of America

412

BWX Technologies Inc., United States of America
Last Energy Inc., United States of America
Westinghouse Electric Company LLC, United States of America
University of California, Berkeley, United States of America
Elysium Industries, United States of America
Oklo Inc., United States of America
HolsGen LLC, United States of America
X-energy, United States of America
General Atomics, United States of America
Ultra Safe Nuclear Corporation, United States of America
ThorCon International, United States of America and Indonesia
StarCore Nuclear, Canada, United Kingdom and United States of America

IAEA Staff

Batra, C.

Division of Nuclear Power, Nuclear Power Technology Development
Section

Cartas, A.R.

Division of Nuclear Power, Nuclear Power Technology Development
Sectiondes Cloizeaux, A.

Division of Nuclear Power

Constantin, A.

Division of Nuclear Power, Nuclear Power Technology Development
SectionDardour, S.

Division of Planning, Information and Knowledge Management, Planningand Economic Studies Section

Ganda, F.

Division of Nuclear Power, Nuclear Power Technology Development
SectionGastl, C.

Division of Nuclear Fuel Cycle and Waste Technology, Nuclear Fuel Cycle
and Materials Section

Gonzalez Espartero,A.

Division of Nuclear Fuel Cycle and Waste Technology, Nuclear Fuel Cycle
and Materials Section

van Heek, A.

Division of Planning, Information and Knowledge Management, Planningand Economic Studies Section

413

Janski, S.

Division of Nuclear Fuel Cycle and Waste Technology, Section on
Decommissioning and Environmental Remediation

Jevremovic, T.

Division of Nuclear Power, Nuclear Power Technology Development
SectionKriventsev, V.

Division of Nuclear Power, Nuclear Power Technology Development
SectionLange, K.

Division of Nuclear Fuel Cycle and Waste Technology, Waste Technology
SectionLepouze, B.

Division of Nuclear Power, Nuclear Power Technology Development
SectionMayer, S.J.

Division of Nuclear Fuel Cycle and Waste Technology, Waste Technology
SectionMichal, V.

Division of Nuclear Fuel Cycle and Waste Technology, Section on
Decommissioning and Environmental Remediation

Monti, S.

Division of Nuclear Power, Nuclear Power Technology Development
Section

Paillere, H.

Division of Planning, Information and Knowledge Management, Planningand Economic Studies Section

Peguero, L.

Division of Nuclear Power, Nuclear Power Technology Development
Section

Pereira, G.

Division of Nuclear Fuel Cycle and Waste Technology, Section on
Decommissioning and Environmental Remediation

Subki, M.H.
(Technical Officer)

Division of Nuclear Power, Nuclear Power Technology Development
SectionZou, Y.
(Technical Officer)

Division of Nuclear Power, Nuclear Power Technology Development
Section414

For further information:
Nuclear Power Technology Development Section (NPTDS)
Division of Nuclear Power
IAEA Department of Nuclear Energy
International Atomic Energy Agency
Vienna International Centre
PO Box 100
1400 Vienna, Austria
Telephone: +43 1 2600-0
Fax: +43 1 2600-7
Email: Official.Mail@iaea.org
Internet: www.iaea.org

22-02924E

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September 2022