spine = 20.5 mm, 80gm paper

Advances in Small Modular Reactor
Technology Developments
Advances in Small Modular Reactor Technology Developments

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: http://www.iaea.org

16-24301

Printed by IAEA in Austria
August 2016

A Supplement to:
IAEA Advanced Reactors Information System (ARIS)
2016 Edition

ADVANCES IN SMALL MODULAR REACTORTECHNOLOGY DEVELOPMENTS
2016 Edition
A Supplement to:
IAEA Advanced Reactors Information System (ARIS)
http://aris.iaea.org

DISCLAIMER
This is not an official IAEA publication. The views expressed do not necessarily reflect those of theInternational Atomic Energy Agency or its Member States.
The material has not undergone an official review by the IAEA. This document should not be quoted
or listed as a reference.
The use of particular designations of countries or territories does not imply any judgement by the
IAEA, as to the legal status of such countries or territories, of their authorities and institutions or ofthe delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated as registered) does
not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or
recommendation on the part of the IAEA.

Printed by the IAEA in Austria
August 2016

FOREWORD
The IAEA Department of Nuclear Energy continues to facilitate efforts of Member States in thedevelopment and deployment of small modular reactors (SMRs), recognizing their potential as a
viable solution to meet energy supply security, both in newcomer and expanding countries interested
in SMRs. In this regards, balanced and objective information to all Member States on technology
status and development trends for advanced reactor lines and their applications are collected, assessedand 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 advanced SMR designs
and concepts, as well as current development trends. The IAEA Division of Nuclear Power, which has
been facilitating Member States in addressing common technologies and issues for SMRs and therelated fuel-cycle, plays a prominent role in convening international scientific forums and technicalcooperation in this field for the interested Member States. The activities on SMRs are further
supported by specific activities on advance fast and high temperature gas cooled reactor technology
development.
The driving forces in the development of SMRs are their specific characteristics. They can be
deployed incrementally to closely match increasing energy demand resulting in a moderate financial
commitment for countries or regions with smaller electricity grids. SMRs show the promise ofsignificant 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 suited for partial or dedicated use in non-electrical applications such as providing
heat for industrial processes, hydrogen production or sea-water desalination. Process heat orcogeneration results in significantly improved thermal efficiencies leading to a better return on
investment. Some SMR designs may also serve niche markets, for example to burn nuclear waste.
Booklets on the status of SMR technology developments have been regularly published. 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 water-cooled reactors, hightemperature gas cooled reactors, liquid metal, sodium and gas-cooled fast neutron spectrum reactors
and molten salt reactors.
This booklet is intended as a supplement to the IAEA Advanced Reactor Information System (ARIS),
which can be accessed at http://aris.iaea.org. Other recent booklets published in support of ARIS arelisted in Annex V.
This publication was developed by Nuclear Power Technology Development Section, Division of
Nuclear Power of the IAEA Department of Nuclear Energy in cooperation with Member States. The
IAEA officers responsible for this publication were S. Banoori, M. H. Subki and F. Reitsma of theDivision of Nuclear Power.

TABLE OF CONTENTSINTRODUCTION .................................................................................................................................. 1
WATER COOLED SMALL MODULAR REACTORS (LAND BASED) ......................................... 11
CAREM (CNEA, Argentina) ........................................................................................................ 13
ACP100 (CNNC, China)............................................................................................................... 21
CAP150 (SNERDI/SNPTC, China) .............................................................................................. 29
CAP200 (SNERDI/SNPTC, China) .............................................................................................. 37
AHWR-300 (BARC, India) .......................................................................................................... 45
IRIS (IRIS International Consortium)........................................................................................... 53
DMS (Hitachi-GE Nuclear Energy, Japan) ................................................................................... 61
IMR (Mitsubishi Heavy Industries, Japan) ................................................................................... 69
SMART (KAERI, Republic of Korea) .......................................................................................... 77
UNITHERM (NIKIET, Russian Federation) ................................................................................ 85
KARAT-45 (NIKIET, Russian Federation)................................................................................... 92
KARAT-100 (NIKIET, Russian Federation)................................................................................. 99
ELENA (Kurchatov Institute, Russian Federation) .................................................................... 106
RUTA-70 (NIKIET, Russian Federation) ................................................................................... 113
NuScale (NuScale Power Inc., United States of America) ......................................................... 121
mPower (BWX Technologies, Inc., United States of America) .................................................. 129
Westinghouse SMR (Westinghouse Electric Company, LLC., United States of America)......... 137
SMR-160 (Holtec International, United States of America) ...................................................... 145
WATER COOLED SMALL MODULAR REACTORS (MARINE BASED) .................................. 153
ACPR50S (CGN, China) ............................................................................................................ 155
Flexblue (DCNS, France) ........................................................................................................... 163
KLT-40S (OKBM Afrikantov, Russian Federation) ................................................................... 170
RITM-200 (OKBM Afrikantov, Russian Federation) ................................................................. 178
VBER-300 (OKBM Afrikantov, Russian Federation) ................................................................ 186
ABV-6E (OKBM Afrikantov, Russian Federation) .................................................................... 194
SHELF (NIKIET, Russian Federation) ....................................................................................... 202
HIGH TEMPERATURE GAS COOLED SMALL MODULAR REACTORS.................................. 211
HTR-PM (Tsinghua University, China) ...................................................................................... 213
GTHTR300 (Japan Atomic Energy Agency, Japan) ................................................................... 221
GT-MHR (OKBM Afrikantov, Russian Federation) .................................................................. 229
MHR-T reactor/Hydrogen production complex (OKBM Afrikantov, Russian Federation) ....... 234
MHR-100 (OKBM Afrikantov, Russian Federation) .................................................................. 239
PBMR-400 (Pebble Bed Modular Reactor SOC Ltd., South Africa).......................................... 245
HTMR-100 SMR (Steenkampskraal Thorium Limited (STL), South Africa ............................. 251
SC-HTGR (AREVA NP, USA) ................................................................................................... 259
Xe-100 (X-energy, United States of America) ............................................................................ 267

FAST NEUTRON SPECTRUM SMALL MODULAR REACTORS ................................................ 273
LEADIR-PS (Northern Nuclear Industries Incorporated, Canada) ............................................ 275
4S (Toshiba Corporation, Japan) ................................................................................................. 283
BREST-OD-300 (NIKIET, Russian Federation) ........................................................................ 291
SVBR-100 (JSC AKME Engineering, Russian Federation) ...................................................... 299
G4M (Gen4 Energy Inc., United States of America) .................................................................. 307
EM2 (General Atomics, United States of America)..................................................................... 314
MOLTEN SALT SMALL MODULAR REACTORS ........................................................................ 321
Integral Molten Salt Reactor (Terrestrial Energy, Canada) ......................................................... 323
MSTW (Seaborg Technologies, Denmark) ................................................................................. 331
ThorCon (Martingale, International Consortium) ....................................................................... 337
FUJI (International Thorium Molten-Salt Forum, Japan) ........................................................... 345
Stable Salt Reactor (Moltex Energy, United Kingdom) ............................................................. 353
SmAHTR (Oak Ridge National Laboratory, United States of America) .................................... 361
Liquid Fluoride Thorium Reactor (Flibe Energy, United States of America) ............................. 369
Mk1 PB-FHR (UC Berkeley, United States) .............................................................................. 377
ANNEX I SMR Estimated Timeline of Deployment ......................................................................... 385
ANNEX II Summary of SMR Designs Based on Power Range ........................................................ 386
ANNEX III Summary of SMR Designs Based on Core Exit Temperature ....................................... 387
ANNEX IV Summary of SMR Designs for Non-Electric Applications ............................................ 388
ANNEX V Bibliography .................................................................................................................... 389
ANNEX VI Acronyms ....................................................................................................................... 391

INTRODUCTION
The IAEA’s Department of Nuclear Energy within its structure contains the Section for
Nuclear Power Technology Development that is tasked to facilitate efforts of Member Statesin identifying key enabling technologies in the development of advanced reactor lines andaddressing their key challenges in near term deployment. By establishing international
networks and ensuring coordination of Member State experts, publications on international
recommendations and guidance focusing on specific needs of newcomer countries are issued.
There is increasing interest in small modular reactors (SMRs) and their applications. SMRs
are newer generation reactors designed to generate electric power up to 300 MW, whose
components and systems can be shop fabricated and then transported as modules to the sites
for installation as demand arises. Most of the SMR designs adopt advanced or even inherent
safety features and are deployable either as a single or multi-module plant. SMRs are underdevelopment for all principal reactor lines: water cooled reactors, high temperature gas
cooled reactors, liquid-metal, sodium and gas-cooled reactors with fast neutron spectrum, andmolten salt reactors. The key driving forces of SMR development are fulfilling the need forflexible power generation for a wider range of users and applications, replacing ageing fossilfired units, enhancing safety performance, and offering better economic affordability.
Many SMRs are envisioned for niche electricity or energy markets where large reactors
would not be viable. SMRs could fulfil the need of flexible power generation for a wider
range of users and applications, including replacing aging fossil power plants, providing
cogeneration for developing countries with small electricity grids, remote and off grid areas,and enabling hybrid nuclear/renewables energy systems. Through modularization technology,
SMRs target the economics of serial production with shorter construction time. Near term
deployable SMRs will have safety performance comparable or better to that of evolutionary
reactor designs.
Though significant advancements have been made in various SMR technologies in recent
years, some technical issues still attract considerable attention in the industry. These include
for example control room staffing and human factor engineering for multi-module SMR
plants, defining the source term for multi-module SMR plants with regards to determining the
emergency planning zone, developing new codes and standards, and also load-following
operability aspects. Some potential advantages of SMRs like the elimination of public
evacuation during an accident or a single operator for multiple modules are being challenged
by regulators. Furthermore, although SMRs have lower upfront capital cost per unit, theirgenerating cost of electricity will probably be substantially higher than that for large reactors.
Currently there are at least 50 SMR designs under development for different application.
Three industrial demonstration SMRs are in advanced stage of construction: in Argentina(CAREM, an integral PWR), in China (HTR-PM, a high temperature gas cooled reactor) andin the Russian Federation (KLT40s, a floating power unit). They are scheduled to start
operation between 2017 and 2020. Dozens of SMR designs are also being prepared for nearterm deployment. Given the above mentioned issues and challenges to be resolved,
realistically the first commercial fleet of SMRs is expected to start up in the time frame of2025 – 2030. However, large fleet deployment of SMRs will only take place beyond 2030.
To assist the reader to easily understand the current status of deployment Table 1 lists all the
SMR designs with the applicable technology along with the output capacity, type of reactor
and design institute information.
1

Table 1: Status of Deployment of SMR Designs and Technologies
DesignKLT-40S

Output70

Type
Floating NPP

Designers
OKBM Afrikantov

Country
Russian Federation

HTR-PM

210

HTGR

INET, Tsinghua University

China

CAREM

30

PWR

CNEA

Argentina

ACP100

100

PWR

CNNC

China

CAP150/200

150/200

PWR

CGNPC

China

AHWR-300

300

PHWR

BARC

India

IRIS

335

PWR

IRIS Consortium

Multi Countries

DMS

300

BWR

Hitachi GE

Japan

IMR

350

PWR

MHI

Japan

SMART
UNITHERM

100
6.6

PWR
PWR

KAERI
NIKIET

Republic of Korea
Russian Federation

KARAT-45/100

45/100BWRNIKIET

Russian Federation

ELENA

68 kW

PWR

RUTA-70PWR

NIKIET

NuScale70
MW(th)
50 × 12PWR

NuScale PowermPower

195 × 2PWR

BWX Technologies

W-SMR

225

PWR

Westinghouse

SMR-160

160

PWR

Holtec International

ACPR50S

60

PWR

CGNPC

United States ofAmerica
China

Flexblue

165

DCNS

France

RITM-200
VBER-300
ABV-6E
SHELF

50 × 2
325
6
6.4

OKBM Afrikantov
OKBM Afrikantov
OKBM Afrikantov
NIKIET

Russian Federation
Russian Federation
Russian Federation
Russian Federation

GTHTR300
GT-MHR

300
285

Immersed
NPP
Floating NPP
Floating NPP
Floating NPP
Immersed
NPP
HTGR
HTGR

JAEA
OKBM Afrikantov

Japan
Russian Federation

MHR-T

205.5х4

HTGR

OKBM Afrikantov

Russian Federation

MHR-100

25 – 87

HTGR

OKBM Afrikantov

Russian Federation

PBMR-400

165

HTGR

PBMR SOC LtdSouth Africa

HTMR-10035

HTGR

Steenkampskraal Thorium
Limited (STL)South Africa

2

Russian Federation
United States ofAmerica
United States ofAmerica
United States ofAmerica
United States ofAmerica

StatusUnder
construction
Under
construction
Under
construction
Conceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignCertified Design
Conceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignUnder
Development
Under
Development
Conceptual
designcompleted
Conceptual
DesignConceptual
Designcompleted
Conceptual
DesignDetailed Design
Licensing Stage
Final design
Detailed Design
Underway
Basic Design
Preliminary
Designcompleted
Conceptual
DesignConceptual
DesignPreliminary
Designcompleted
Advanced
Conceptual
Design phaseDesign
SC-HTGR

Output
272

Type
HTGR

Designers
AREVA

Xe-10035

HTGR

X-energy LLCLEADIR-PS

39

LMFR

4S
BREST-OD-300
SVBR-100
G4M

10
300
100
25

LMFR
LMFR
LMFR
LMFR

Northern Nuclear
Industries Incorporated
Toshiba Corporation
NIKIET
JSC AKME Engineering
Gen4 Energy Inc.

EM2

265

GMFR

General Atomics

IMSR

MSR

Terrestrial Energy

MSTW

185192
100

Japan
Russian Federation
Russian Federation
United States ofAmerica
United States ofAmerica
CanadaMSR

Seaborg Technologies

Denmark

ThorCon

250

MSR

Martingale

FUJI

200

MSR

Stable Salt Reactor

37.5×8MSRInternational Thorium
Molten-Salt Forum:
ITMSF
Moltex Energy

International
ConsortiumJapan

SmAHTR

125
MW(th)
250

MSR

ORNL

MSR

Flibe Energy

100

MSR

University of California,
Berkeley

LFTR
Mk1 PB-FHR

Country
United States ofAmerica
United States ofAmerica
CanadaUnited Kingdom
United States ofAmerica
United States ofAmerica
United States ofAmerica

StatusConceptual
DesignConceptual
DesignConceptual
DesignDetailed Design
Detailed Design
Detailed Design
Conceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignConceptual
DesignCompleted
Conceptual
DesignPre-conceptual
designConceptual
DesignPre-Conceptual
DesignFigure 1 shows a pictorial view of all the Member States with Small Modular Reactors
designs and technologies included in the booklet.
Additional summaries are also provided. Annex I provides the estimated timeline of thepossible commercial deployment for the different SMR designs. This is based on the current
design and licensing status but is of course also dependent on project financing and otherfactors. The reactors under construction have the highest probability for earlier deployment.
Annex II provides a summary as a function of different power ranges while Annex III shows
SMR designs based on the core exit coolant temperature. This may be helpful to selectdesigns for specific power or process heat applications, further illustrated in Annex IV for
different non-electric applications.
This booklet is a supplement to the IAEA Advanced Reactor Information System (ARIS,
http://aris.iaea.org). Other recent booklets published in support of ARIS are listed in Annex V.
Annex VI contains a list of commonly used acronyms.

3

Figure 1: Member States with Small Modular Reactors
This booklet provides a brief introductory information and technical description of the key
SMR designs and technologies under different stages of development and deployment. The
2016 edition of the booklet includes the SMR designs with a fast neutron spectrum and forthe first time molten salt reactors (MSRs) designs have also been added. This booklet
comprises five (5) parts arranged in the order of the different types of coolants and neutron
spectrum adopted:
Part One (Land-based water-cooled SMRs) presents the key SMR designs adopting
integral light water reactor (LWR) technologies.
The Central Argentina de Elementos Modulares (CAREM) reactor under development is asmall, integral type pressurized light water reactor (PWR) design, with all primary
components located inside the reactor vessel and an electrical output of 150–300 MW(e). Site
excavation work for a 27 MW(e) CAREM-25 prototype was completed at the end of August2012 and construction formally began in February 2014. In July 2012, the Korean Nuclear
Safety and Security Commission issued the Standard Design Approval for the 100 MW(e)System Integrated Modular Advanced Reactor (SMART) – the first integrated PWR toreceive certification. In September 2015, a pre-project engineering agreement was signed forthe deployment of a SMART reactor in Saudi Arabia.
The China National Nuclear Corporation (CNNC) is developing the ACP100 design and has
undertaken the IAEA Generic Reactor Safety Review (GRSR) in 2015. An industrial

4

demonstration plant with two modules of ACP100 is planned to be deployed in Fujian
Province. The Shanghai Nuclear Engineering Research and Design Institute (SNERDI), a
subsidiary of the State Nuclear Power Technology Corporation (SNPTC), is developing the
China Advanced Passive pressurized water reactors CAP150 and CAP200 The concept
designs adopted passive engineered safety features and simplified systems compared tocurrent operating PWRs, to generate electric power of 150 MW(e) and 200MW(e)respectively.
For heavy-water cooled SMRs, India has developed the advanced heavy water reactor(AHWR300-LEU) to generate 304 MW(e). The design incorporates vertical pressure tubes,
low enriched uranium and thorium fuel, and passive safety features. The basic design has
been completed. The IRIS international consortium has completed the basic design of the
IRIS 325 MW(e) integral PWR. In Japan the DMS design (Double MS: Modular Simplified
and Medium Small Reactor) is a small-sized boiling water reactor (BWR) which generates
300 MW(e). The Integrated Modular Water Reactor (IMR) is a medium sized power reactor
producing electricity of 350 MW(e). Validation testing, research and development forcomponents and design methods, and basic design development are required before licensing.
In the Russian Federation the N.A. Dollezhal Research and Development Institute of PowerEngineering (NIKIET) is designing the UNITHERM to generate 6.6 MW(e), based on
design experience in marine nuclear installations. NIKIET is also working on KARAT-45, a
small boiling water reactor (BWR) with a rated power of 45 MW; KARAT-100, an integral
type multi-purpose BWR with a power output of 360 MW(th) and a rated electrical output of
100 MW; and RUTA-70 a water-cooled water-moderated integral pool-type heating reactor
with a thermal capacity of 70 MW(th) dedicated for district heating or potable waterproduction. The Kurchatov Institute is proposing the ELENA conceptual design, a nuclear
thermoelectric plant to supply 68 kW(e) of electricity and 3.3 MW(th) of heating capacity for
25 year without refuelling.
Four integral pressurized water SMRs are under development in the USA: NuScale, mPower,
the Westinghouse SMR and SMR-160. NuScale is a nuclear power plant made up of twelve
modules each producing 50 MW(e) and design certification application is expected to besubmitted to the US NRC by end of 2016, followed by application for a combined
construction and operation licence (COL). Its design certification target is 2020, for the firstplant that is to be built in Idaho. The mPower design consists of two 195 MW(e) modules.
The Westinghouse SMR is a conceptual design with an electrical output of 225 MW(e),incorporating passive safety systems and proven components of the AP1000. The SMR-160
design generates power of 160 MW(e) adopting passive safety features and its preliminary
design is expected to be completed by the end of 2017.
Part Two (Marine-based water-cooled SMRs) discusses concepts such as the KLT-40S(under construction), ACPR50S, and Flexblue.
In the Russian Federation, a barge mounted floating nuclear power plant (NPP) with two
modules of KLT-40S, a compact PWR with a capacity of 35 MW(e) per module and used forcogeneration of process heat and electricity, is under construction and its commercial startup
is expected in 2019. The ACPR50S is a small modular offshore floating reactor developed by
the China General Nuclear Power Group (CGNPC). It is intended as a potential optimal
solution for combined supply of heat, electricity and fresh water for marine resource
development activities, energy supply and emergency support on islands and along the5coastal area. It is currently in a preliminary design stage with commissioning and connection
to grid expected in 2020. In France, the Flexblue design is a small seabed nuclear power
plant with an output of 160 MW(e). Technical discussions have been initiated with the Frenchtechnical safety authority.
In the Russian Federation the ABV-6E is being developed by OKBM Afrikantov with anelectrical output of 6 MW(e) in cogeneration mode. It is a nuclear steam generating plantusing natural circulation for its integral reactor coolant system. Its development is at the final
design stage. The RITM-200, an integral reactor with forced circulation for multipurpose
nuclear icebreaker, floating or land-based NPP, is designed to generate 50 MW(e). Two serial
RITM-200 icebreakers will be commissioned in 2020. 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 N.A. Dollezhal Research and Development Institute of PowerEngineering (NIKIET) is designing the SHELF, a 6 MW(e) underwater, remotely operated
power source.
Part Three (High Temperature Gas Cooled SMRs) provides information on the HTR-PM
design under construction, other HTGR designs under development for potential certification,and details on the fundamental safety and design strategies.
In the area of innovative reactors, high temperature gas cooled reactors (HTGRs) have an
inherent safety benefit whereby extremely high temperature values can be tolerated without
fuel damage, and this also provides high temperature heat (≥750°C) that can be utilized formore efficient electricity generation, a variety of industrial applications as well as forcogeneration. This enables HTGRs to contribute to the total energy market in addition toelectricity generation. As generally applied for water-cooled SMRs, the smaller size andsimplified design (with a small number of safety systems) also make HTGRs potentiallyattractive to Member States with small electricity grids. Furthermore, the significant process
heat options far exceed the capability of light water reactors and hence make them even more
attractive.
The construction of the High Temperature Reactor–Pebble Bed Module (HTR-PM)
industrial demonstration power plant in China will also make such technology available fornear term deployment. The HTR-PM is a unique twin nuclear steam supply system feeding asingle 200 MW(e) superheated steam turbine generator. Construction has started in December
2012 and the major equipment (RPV, core barrel etc.) have recently been delivered at site,with operation expected by the end of 2017.
Japan Atomic Energy Agency (JAEA) has several conceptual designs based on the experienceand development work related to the HTTR test reactor. One of these designs is the
GTHTR300 (Gas Turbine High Temperature Reactor 300 MW(e)), a multipurpose,
inherently-safe and site-flexible small modular reactor. This reactor is under development forcommercialization in the 2020s. The development of HTGR technologies in Russia includes
the project to develop the 285 MW(e) Gas Turbine-Modular Helium Reactor (GT-MHR) for
electricity production but recently also the diversification of the nuclear power applications
for industrial purposes, such as the MHR-T reactor/hydrogen production complex that use
4x600 MW(th) modules. Finally, the MHR-100 prismatic modular helium reactor design of
215 MW(th) is used in multiple configurations for electricity and cogeneration. The GTMHR is a Russian Federation – USA jointly funded project, originally aimed at solving one

6

of the most important tasks in the area of non-proliferation; the disposition of weapons-grade
plutonium.
The South African developed Pebble Bed Modular Reactor (PBMR-400) design can produce
electricity of 165 MW(e) at high efficiency via a direct Brayton cycle employing a helium gas
turbine. The unique fixed central column design allows the larger thermal power in a pebble
bed design while retaining its inherent safety characteristics for decay heat removal with
passive only means even under the most severe conditions. The project was stopped in 2010
but the design information is secured and maintained. Also in South Africa the
Steenkampskraal Thorium Limited (STL) company is aiming to finish the conceptual designof the HTMR-100 pebble bed design by 2018 and plans to be able to use a range of uranium
and plutonium-thorium coated particle pebble fuels in a once through-then-out (single pass)cycle. The HTMR-100 is designed to generate 35 MW(e).
In the USA, the NGNP Industry alliance has selected the AREVA 272 MW(e) SC–HTGR aprismatic block design for commercialization and the preparation for pre-licensing
application have started. The design is based on the AREVA’s ANTARES concept but
coupled to two steam generators and allows for cogeneration. A privately owned and funded
initiative in the USA, called X-energy, is pursuing the Xe-100 small-sized pebble bed reactor
producing 35 MW(e). A major aim of the design is to improve the economics through system
simplification, component modularization, reduction of construction time and high plant
availability brought about by continuous fuelling.
Part Four (Fast Neutron Spectrum SMRs) presents the SMRs with fast neutron spectrumwith all the different coolant options.
In the booklet on “Status of Innovative Fast Reactor Designs and Concepts” (see Annex V)
the four major fast reactor options were described. They are sodium cooled fast reactor(SFR), the heavy liquid metal-cooled (HLMC) fast reactor, the gas-cooled fast reactor (GFR)
and molten salt fast reactor (MSFR). In this booklet only fast reactor designs with the first
three types of coolant are included. The MSR designs, all with thermal neutron spectrum, areincluded in part five.
The Lead-cooled Integral Reactor-Passively Safe (LEADIR-PS), a small nuclear reactor
being developed by the Northern Nuclear Industries Incorporated (N2I2) is intended toprovide diverse electricity and process heat demands of Canada. LEADIR-PS plants underdevelopment are LEADIR-PS30 (30 MW(th), up to 11 MW(e)), LEADIR-PS100 (100MW(th), up to 39 MW(e), and LEADIR-PS300 (300 MW(th), up to 120 MW(e)). The 4S(super-safe, small and simple) is a small sodium cooled fast reactor without on-site refuelling.
Being developed by Toshiba Corporation, Japan as distributed energy source for multipurpose applications, the 4S offers two power outputs of 30 MW(th) and 135 MW(th),respectively.
In the Russian Federation the N.A. Dollezhal Research and Development Institute of PowerEngineering (NIKIET) is designing the BREST-OD-300 lead cooled fast reactor that uses a
two circuit heat transport system to deliver heat to a supercritical steam turbine and generate
electricity of 300 MW. The BREST-OD-300 power unit is designed as a pilot anddemonstration unit and is considered the prototype of a fleet of medium sized lead cooled fast
reactors. The SVBR-100 is a multipurpose small modular fast reactor lead–bismuth (LBE)
cooled being developed by JSC AKME Engineering, Russian Federation, with an equivalent7electric power of 100 MW. Physical and power generation start-up of its pilot plant is
expected in 2019.
From the United States of America two fast reactor designs are included. The G4M
conceptual design is being developed by the Gen4 Module (G4M) Inc. It is a low capacity,
portable and self-contained reactor designed to generate 25 MW(e). It uses lead–bismuth
eutectic as the primary coolant. The G4M does not need a human operator and employs a fuelcycle of 10 years with no on-site refuelling. Niche markets include producing steam/electricity for mining/oil and gas production sites, isolated and island communities andgovernment facilities. The Energy Multiplier Module (EM2) is a helium-cooled fast reactor
developed by General Atomics. It is designed as a modular, grid-capable power source with a
net unit output of 265 MW(e). With a core outlet temperature of 850°C, it can also be used asa high temperature process heat source. The reactor employs a “convert and burn” core
design which converts fertile isotopes to fissile and burns them in situ over a 30-year core
life.
Part Five (Molten Salt SMRs) presents the SMRs that utilize molten salt fuelled (andcooled) advanced reactor technology. Molten salt reactors (MSRs) have several benefits andinclude (somewhat dependent on the specific design, and relative to currently deployed watercooled reactors: (i) higher operating temperatures leading to increased overall efficiencies;(ii) low coolant pressure; (iii) reduced volume and lifetime of high level waste; (iv) salient
safety characteristics; (v) in the case of MSRs with molten fuel, elimination of technicalchallenges associated with solid fuel related to high burn-up effects; and (vi) flexibility offuel cycles (uranium, plutonium, thorium).
The Integral Molten Salt Reactor – 400 Megawatts-thermal (IMSR400) is a small modular
molten salt fuelled reactor developed by Terrestrial Energy Canada. It electrical capacity is~190 MW(e). All key components of IMSR are permanently sealed during operation andreplaced completely, as a single unit, at the end of its useful service life (nominally 7 years).
Pre-licensing vendor design review of the IMSR is currently underway. The Seaborg
Technologies’ Molten Salt Thermal Wasteburner (MSTW) is a concept with a single molten
fuel-salt to be operated on a combination of spent nuclear fuel and thorium. The design
efforts are currently focused on advanced multi-physics, neutronics, and early engineering
details.
ThorCon is a thermal spectrum molten salt nuclear reactor that utilizes 250 MW(e) power
modules. Each module contains two replaceable reactors in sealed Cans . When one reactor is
producing power the other is in cooldown mode. Cans are replaced every four years when thefuelsalt is transferred to the new Can. The low pressure and short 4-year Can plant life means
ThorCon plans to use normal piping thicknesses and easily automated, ship-style steel plate
construction methods to reduce costs. Currently a pre-feasibility study is being done inIndonesia. Under development in Japan, through the International Thorium Molten-Salt
Forum, the FUJI concept design can be used for transmutation of Pu and/or minor actinides.
It aims not only to generate electricity of 200 MW(e) but also heat for hydrogen production(it has a high exit temperature of 704 °C) and water supply by desalination of seawater.
The Stable Salt Reactor (SSR), being developed by Moltex Energy in the United Kingdom is
unique in its use of molten salt fuel replacing solid pellets in conventional solid fuelassemblies. This brings the major advantages of safe molten salts without the technicalhurdles of managing a mobile liquid fuel. The proposed power plant will deliver 300MWe8made up by eight 37.5MWe modules. The conceptual design has been completed and the key
claims have been independently validated.
The MSR concept was originally developed in the USA headed by the Oak Ridge National
Laboratory (ORNL). The ORNL program culminated in the construction and successful
operation of the Molten Salt Reactor Experiment (MSRE). The MSRE employed a molten
fluoride fuel salt, which was also the primary coolant. Three MSR designs from the USA are
included.
SmAHTR is a small 125 MW(th) fluoride salt–cooled high-temperature reactor (FHR)
design concept being developed by ORNL intended to match the energy requirements ofcoupled industrial processes (for efficient hydrogen production as example). It is not intended
for electricity generation. The purpose is to enable an integrated assessment of the potential
performance of deliberately small FHR designs and to provide guidance to the overall FHR
research and development effort.
The liquid-fluoride thorium reactor (LFTR) design by Flibe Energy is a graphite-moderated,thermal-spectrum reactor with solutions of liquid fluoride salts containing both fissile andfertile materials. The fuel and blanket salts are kept separated in two plena integrated into asingle structure within the reactor vessel. A closed-cycle gas turbine power conversion system
is proposed for 250 MW(e) electrical generation with the objective to produce electricity at
low cost by efficiently consuming thorium. The Mark 1 Pebble-Bed Fluoride-Salt-Cooled
High-Temperature-Reactor (Mk1 PB-FHR) is a small, modular graphite-moderated reactor.
It is being developed by University of California, Berkeley. The design is differentiated fromother MSRs because it use coated particle fuels, cooled by the fluoride salt flibe. The power
plant 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 peakelectricity generation.
In this booklet, effort has been made to present all SMR designs within the above five parts.
Each description includes a general design description and philosophy, target applications,development milestone, nuclear steam supply system, a comprehensive table of the major
design parameters, and then descriptions of the reactor core, engineered safety features, plantarrangement, design and licensing status, and finally plant economics. Not all small reactor
designs presented can strictly be categorized as small modular reactors. Some strongly rely
on proven technologies of operating large capacity reactors, while others do not use amodular or integral design approach. They are presented in this booklet for reason ofcompleteness and since designers foresee certain niche markets for their products.
The technical description and major technical parameters, including the predictions of coredamage frequencies (CDFs), were provided by the design organizations without validation orverification by the IAEA. All figures, illustrations and diagrams were also provided by thedesign organizations.
It is hoped that this Booklet will be useful to Member States with a general interest in SMRs,
as well as to those newcomer countries looking for more specific technical information. Itshould also further promote contributions to and the use of the IAEA Advanced Reactor
Information System (ARIS).

9

10

WATER COOLEDSMALL MODULAR REACTORS(LAND BASED)

12

CAREM (CNEA, Argentina)
1. Introduction
CAREM is a national SMR development project based on LWR technology coordinated by
the Argentina’s National Atomic Energy Commission (CNEA) in collaboration with leading
nuclear companies in Argentina with the purpose to develop, design and construct innovativesmall nuclear power plants with high economic competitiveness and level of safety. CAREM
is an integral type PWR based on indirect steam cycle with distinctive features that simplify
the design and support the objective of achieving a higher level of safety. CAREM reactor
was developed using domestic technology, at least 70% of the components and related
services for CAREM were sourced from Argentinean companies. The prototype module of
CAREM (CAREM-25) has a thermal power of 100 MW (30 MW(e)) is currently underconstruction.

Figure 1: CAREM module with its second shutdown and residual heat removal systems(Reproduced courtesy of CNEA)
2. Target Applications
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 tocoastal sites.
3. Development Milestones
1984

CAREM concept was presented in Lima, Peru, during the IAEA Conference on SMRs and was one of thefirst of the new generation reactor designs. CNEA officially launched the CAREM project

2001-02

The design was evaluated on Generation IV International Forum and was selected in the near term
development group

2006

Argentina Nuclear Reactivation Plan listed the CAREM-25 project among priorities of national nuclear
development
CNEA submitted its preliminary Safety Analysis Report (PSAR) for CAREM-25 to the ARN.
Announcement was made that Formosa province was selected to host the CAREM

2009

13

2011

Start-up of a high pressure and high temperature loop for testing the innovative hydraulic control rod drive
mechanism (CAPEM)

2011

Site excavation work began and contracts and agreements between stakeholders are under discussion

2012

Civil engineering works

2014

8 February, formal start of construction

2019

First fuel load

4. General Design Description
Design Philosophy
CAREM is a natural circulation based indirect-cycle reactor with distinctive features that
simplify the design and improving safety performance. Its primary circuit is fully containedin the reactor vessel and it does not need any primary recirculation pumps. The pressurizationis achieved by balancing vapour production and condensation in the vessel. The CAREM
design reduces the number of sensitive components and potentially risky interactions with theenvironment.
Some of the significant design characteristics are:• integrated primary cooling system;• self-pressurized;• core cooling by natural circulation;• in-vessel control rod drive mechanisms;• safety systems relying on passive features;• Balanced and optimized design with a cost-effective internalization of safety.
Nuclear Steam Supply System
CAREM is an integrated reactor. Its high-energy primary system (core, steam generators,
primary coolant and steam dome) is contained inside a single pressure vessel. Primarycooling flow is achieved by natural circulation, which is induced by placing the steam
generators above the core as shown in Figure 1. Water enters the core from the lower
plenum. After being heated, the coolant exits the core and flows up through the chimney tothe upper steam dome. In the upper part, water leaves the chimney through lateral windows to
the external region. It then flows down through modular steam generators, decreasing its
enthalpy. The coolant exits the steam generators and flows through the downcomer to thelower plenum, closing the circuit. At steady-state conditions, the driving forces created by thedensity differences along the circuit are balanced by the friction and form losses, so there isan adequate flow rate in the core. The coolant also acts as a neutron moderator.
Reactor Core
The reactor core of CAREM-25 has fuel assemblies of hexagonal cross section. There are 61
fuel assemblies having about 1.4 meters active length. Each fuel assembly contains 108 fuelrods with 9 mm outer diameter, 18 guide thimbles and one instrumentation thimble as shownin Figure. 2. The fuel is 1.8% - 3.1% enriched UO2.The fuel cycle can be tailored to customer
requirements, with a reference design for the prototype of 510 full-power days and 50% ofcore replacement.
Reactivity Control
Core reactivity is controlled by the use of Gd2O3 as burnable poison in specific fuel rods andmovable absorbing elements belonging to the adjustment and control system. Neutron poisonin 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
14(‘spider’), so the whole cluster moves as a single unit. Absorber rods fit into the guide tubes.
The absorbent material is the commonly used Ag-In-Cd alloy. Absorbing elements are used
for reactivity control during normal operation, and to produce a sudden interruption of thenuclear chain reaction when required.

Figure 2: Reactor Core and of Fuel Assembly of CAREM-25(Reproduced courtesy of CNEA)
Reactor Pressure Vessel and Internals
Reactor Pressure Vessel (RPV) of CAREM-25 has height of 11 meters and 3.4 meters
diameter having variable thickness of 13 cm to 20 cm. The RPV is built in forged steel withinternal lining of stainless steel.

Figure 3: CAREM-25 Reactor Vessel and Internal(Reproduced courtesy of CNEA)

15

Reactor Coolant System
Primary cooling in CAREM is achieved using natural circulation. The primary circuit of
CAREM is fully contained in the reactor vessel and it has no cooling pumps. The reactor coreheats reactor coolant causing the coolant to flow upward through the riser. When the heatedreactor coolant exits the riser, it passes over the tubes of the helical coil steam generators,
which act as a heat sink.
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 eventual pressure perturbations. Due to self-pressurization, bulk
temperature at core outlet corresponds to saturation temperature at primary pressure. In thisway, typical heaters present in conventional PWRs are eliminated.
Steam Generator
In CAREM-25, twelve identical mini-helical vertical steam generators, of the once-throughtype are placed equidistant from each other along the inner surface of the RPV. Each of themconsists of a system of 7 coiled piping layers, 52 parallel pipes of 26m length. They are used
to transfer heat from the primary to the secondary circuit, producing superheated dry steam at
4.7 MPa. The secondary system circulates upwards within the tubes, while the primarycoolant moves in counter-current flow. In order to achieve a mostly uniform pressure-loss
and superheating on the secondary side, the length of all tubes is equalized. Due to safety
reasons, the steam generators are designed to withstand the primary pressure withoutpressure in the secondary side and the whole live steam system is designed to withstandprimary pressure up to isolation valves (including the steam outlet/water inlet headers) incase of SG tube breakage.

Figure 4: Layout of CAREM-25 steam generators inside RPV(Reproduced courtesy of CNEA)

16

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

CNEA

Country of origin

Argentina

Reactor type

Integral PWR

Electrical capacity (MW(e))~30Thermal capacity (MW(th))

100

Expected Capacity Factor (%)˃ 90

Design Life (years)

60

2Plant Footprint (m )Not available

Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure (MPa)

12.25
o

Core inlet/exit temperatures ( C)

284 / 326

Main Reactivity Control Mechanism

Control Rod Driving Mechanism (CRDM) only

RPV Height (m)11

RPV Diameter (m)3.2

RPV, internals and SGs Weight (metric ton)

267

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine Cycle

Fuel Type/Assembly Array

UO2 pellet/hexagonal

Fuel Assembly Active Length (m)1.4

Number of Fuel Assemblies

61

Fuel Enrichment

3.1% (prototype)

Fuel Burnup (GWd/ton)

24 (prototype)

Fuel Cycle (months)

14 (prototype)Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

2

Refuelling Outage (days)

not available

Distinguishing features

Core heat removal by natural circulation, Pressure
suppression containment

Modules per Plant

1

Target Construction Duration (months)~36

Seismic design (g)0.25

Predicted core damage frequency (per reactor year)~1E7

Design Status

Under construction (as prototype)

5. Safety Features
Engineered Safety System Configuration and Approach
The safety system of CAREM consists of two reactor protection systems (RPS), two
shutdown systems, passive residual heat removal system (PRHRS), safety anddepressurization valves, low pressure injection system and containment of pressure
suppression type. The two shutdown systems comply with requirements of redundancy,

17

independence, separation and diversification and act automatically. Each can maintain thecore sub-critical in all shut down states; First Shutdown System (FSS) consists of 9 fast
shutdown rods and 16 reactivity adjust and control rods located over the core. They fall bygravity in case of necessity. Second Shutdown System consists of a gravity driven injection
device of borated water (two tanks) at high pressure which actuates automatically when
failure of the FSS is detected. For the grace period of 36 hours, core decay heat removal canensure safe core temperature due to availability of one out of two PRHRS in the case of lossof heat sink or Station Black-out (SBO). In CAREM, SBO is classified as a design basis
event. The PRHRS are heat exchangers formed by parallel horizontal U-tubes (condensers)
coupled to common headers. A set of headers is connected to the RPV steam dome, while
another set (condensate return line) is coupled with the RPV at the inlet of the primary systemside of the SG. Through natural circulation, the design provides core decay heat removal,transferring it to dedicated pools inside the containment and then to the suppression pool.
Two redundant diesels provide emergency supply for active cooling systems for the longterm. Despite the low frequency of a SBO longer than 36 hours, prevention is considered forgrace period prolongation by simple systems supported by fire extinguishing system orexternal pumps and containment protection.
Regarding severe accident mitigation, provisions are considered for Hydrogen control and for
RPV lower head cooling for in-vessel corium retention. Safety classification of systems,
structures and components (SSCs) important to safety is based on identification of low levelsafety functions (LLSF) -derived from the Fundamental Safety Functions- and safetyfunctional groups of SSCs that fulfil those functions. Criteria for safety categories assignation
to LLSF and classes to SSC are obtained from the way the principle of Defence in Depth isinternalized in the design, and probabilistic and deterministic considerations. Three categories
and classes are defined. This methodology, in accordance with IAEA SSG-30, provides aclear assignation of design rules and requirements to systems important to safety and its
SSCs.

Figure 5: Safety Systems of CAREM-25(Reproduced courtesy of CNEA)
Containment System
The cylindrical containment vessel with a pressure suppression pool is a 1.2 m thickreinforced concrete external wall having a stainless steel liner inner surface and withstands
earthquakes of 0.25g. It is designed to sustain the pressure of 0.5 MPa. Ultimate heat sink18

inside the containment during the grace period provides protection for external extreme
events.

PRHRS Pool “A”

Second Shutdown
System “room B”
Dry-wellPressure
suppression pool
RPV

Figure 6: CAREM-25 containment cut-away view
(Reproduced courtesy of CNEA)
6. Plant Safety and Operational Performances
The natural circulation of coolant produces different flow rates in the primary system
according to the power generated or removed. Under different power transients, a selfcorrecting response in the flow rate is obtained. Due to the self-pressurizing of the RPV(steam dome), the system keeps the pressure very close to the saturation pressure. Under all
operating conditions this could prove to be sufficient to guarantee a remarkable stability of
the RPV pressure response. The control system is capable of keeping the reactor pressure
practically at the operating set point through different transients, even in case of powerramps. The negative reactivity feedback coefficients and the large water inventory in the
primary circuit combined with the self-pressurization features make this behaviour possible
with minimum control rod motion. In addition, a condensation coil is included in the steam
dome in order to tune the reactor operational point. Many events that would lead to accidental
conditions are rendered impossible by the innovative design, like large LOCA, LOFA andcontrol rod ejection between others. On the other hand some postulated multiple failure
events are considered in the design basis as ATWS, PRHRS and ancillary diesels failure.
Moreover large primary water/power ratio makes transients manageable after most severe
system or component failure (large thermal inertia). Deterministic and probabilistic safety
analyses were developed; they fulfil regulatory requirements and large margins against
acceptance criteria are observed.

19

7. Instrumentation and Controls Systems
Plant control is performed by a distributed control system, computer based and with highavailability. There 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 Arrangement

Figure 7: CAREM-25 plant layout (Reproduced courtesy of CNEA)
9. Design and Licensing Status
After completion of licensing process, construction of the Prototype started. Contracts withdifferent Argentinean stakeholders for manufacturing of components issues have alreadybeen signed. Environmental Impact Study was approved by the Local Authority. Nonnuclear buildings first concrete was poured in February 2014.

Figure 8: CAREM-25 progressing in construction as of March 2016
(Reproduced courtesy of CNEA)
10. Plant Economics
As CAREM-25 is a prototype, plant economics is not provided.

20

ACP100 (CNNC, China)
1. Introduction
The ACP100 is an integrated PWR design developed by China National Nuclear Corporation(CNNC) to generate an electric power of 100 MW(e). The ACP100 is based on existing PWR
technology adapting passive safety system, which in case of operational transients andpostulated design basis accidents uses natural convection to cool down the reactor. The
ACP100 integrated design of its reactor coolant system (RCS) enables the installation of its
major primary circuit’s components within the reactor pressure vessel (RPV), as shown in
Figure 1. The ACP100 plant design allows a deployment of one to eight module(s) togenerate larger plant output as demands arise. The plant design contains the following
technical features:• An integrated primary system and equipment layout, eliminating the possibility of alarge break loss of coolant accidents (LOCA);• The reactor and spent fuel pool are installed underground, as shown in Figure 2,
providing better protection against external events, and reduced possibility ofradioactive release;• Large primary coolant inventory resulting in large thermal inertia, thus providing
longer response time in the case of transients or accidents;• Passive severe accident prevention and mitigation systems such as containmenthydrogen eliminator, cavity flooding, etc. ensure the integrity of containment.

Figure 1: Reactor System Configuration of ACP100(Reproduced courtesy of CNNC)

21

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. Development Milestones
2011

2013

CNNC signed an agreement with the Zhangzhou municipal government in Fujian
Province to host the first two ACP100 demonstration units
CNNC signed an agreement with the Chenzhou municipal government in Hunan
Province to deploy an ACP100 power plant
Basic design completed

December 2014

Preliminary safety assessment report (PSAR) approved

April 2015

An agreement to conduct a generic safety review for ACP100 was signed between the
IAEA and CNNC

July 2012

4. General Design Description
Design Philosophy
The ACP100 realizes design simplification by integrating the primary cooling system andenhanced safety by means of passive safety systems.
Nuclear Steam Supply System
The integrated nuclear steam supply system (NSSS) design of the ACP100 consists of thereactor core, sixteen (16) once-through steam generator (OTSG) and a pressurizer that are
installed within the a reactor pressure vessel (RPV). The four (4) canned motor pumps are
installed nozzle to nozzle to the RPV.
Reactor Core
The 57 fuel assemblies (FAs) of ACP100 core with total length of 2.15m core have a squared
17x17 configuration. The expected average fuel enrichment is about 2.4 - 4.0%. The reactor
will be able to operate 24 months per fuel cycle. In August 2012, Jianzhong nuclear fuelfabrication plant (subsidiary of CNNC) at Yibin started manufacturing FAs and control rod
samples for ACP100 R&D project at NPIC.
Reactivity Control
The reactivity is controlled by means of control rods, solid burnable poison and soluble boron
dispersed in the primary coolant. There are 25 control rods, with a magnetic force typecontrol 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 generator. The RPV is protected by a safety relief valves against overpressurization in the case of strong differences between core power and the power removed
from the RPV.
Reactor Coolant System
The ACP100 primary cooling mechanism under normal operating condition and shutdown
condition is done by forced circulation. The RCS has been designed to ensure adequatecooling of reactor core under all operational states, and during and following all postulated
off normal conditions. The four (4) reactor coolant pumps (RCP) are mounted into the RPV
through short stub pipes, enabling the elimination of large bore piping. The integral design of
RCS significantly reduces the flow area of postulated small break LOCA.

22

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

CNNC (NPIC/CNPE)

Country of origin

China

Reactor type

Integral PWR

Electrical capacity (MW(e))

100Thermal capacity (MW(th))

310

Expected Capacity Factor (%)

95

Design Life (years)

60

2Plant Footprint (m )200000

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)

15.00
o

Core inlet/exit temperatures ( C)

282.6/ 323.4

Main Reactivity Control Mechanism

Control rod drive mechanism (CRDM), solid burnable poison andboron solution

RPV Height (m)10

RPV Diameter (m)3.19

Module Weight (metric ton)

250

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine cycle

Fuel Type/Assembly Array

UO2/17x17 square pitch arrangement

Fuel Assembly Active Length (m)2.15

Number of Fuel Assemblies

57

Fuel Enrichment (%)

2.40 – 4.00

Fuel Burnup (GWd/ton)< 45000

Fuel Cycle (months)

24

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

2

Refueling Outage (days)

40

Distinguishing features
Modules per Plant

Integrated reactor with tube-in-tube once through steam generator,
nuclear island under the ground level
1-8

Target Construction Duration (months)

36

Seismic design (g)0.30

Predicted core damage frequency (per reactor year)< 1E–6

Design Status

Basic design

Steam Generator
There are sixteen (16) OTSGs, which are integrated within the RPV. All of 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 the hole on barrel supporting hub, the heads are welded to thereactor vessel vapour cavity.

23

Pressurizer
The pressurizer of ACP100 is located outside of the reactor vessel. The pressurizer is avertical, cylindrical vessel with hemispherical top and bottom heads. Electrical heaters are
installed through the bottom head of the vessel, while the spray nozzle, the spring-operated
safety valves and the automatic depressurization systems (1 and 2 stages) valves connection
nozzles are located in the top head of the vessel.

Figure 2: ACP100 Reactor Building underground location(Reproduced courtesy of CNNC)
5. Safety Features
The ACP100 is designed with inherent and passive safety features, eliminating large bore
primary coolant piping which in turn eliminates large break LOCA. The passive safetysystem mainly consists of the passive decay heat removal system (PDHRS), passive safety
injection system, passive containment heat removal system and reactor depressurization
system (RDP).
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 installing
the reactor building and spent fuel pool below ground level. When the spent fuel pool is filledwith spent fuel of ten (10) years, the cooling system is able to cope for seven (7) days ofcooling in the case of accident before boiling dry and uncovering fuel. Severe accident
prevention and mitigation are achieved through: passive reactor cavity flooding prevents
RPV melt, passive hydrogen recombination system prevents containment hydrogen explosion
and maintain the containment integrity under severe accidents, automatic pressure relief
system and RPV off-gas system to remove non-condensable gas gathered at RPV head after
accidents.
24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., toprevent the change of beyond DBA to severe phase). The PDHRS of the ACP100 consists of
one emergency cooler and associated valves, piping, and instrumentation. The emergency
cooler is located in the containment cooling water storage tank, which provides the heat sink
for the emergency cooler. The PDHRS heat exchanger is installed in the in-containment water
storage tank filled with coolant. The decay heat is removed from the core by natural
circulation or forced circulation by the RCPs. The PDHRS provides cooling for 3 days
without operator intervention or 14 days with water supply from the cooling pool by drainedby gravity force.
Emergency core cooling system
The emergency core cooling system (ECCS) consists of two coolant storage tanks, two
injection tanks, an in-containment water storage tank 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 toseven (7) days. Under accident condition, heat is being removed by gas and steam convection
between containment and ultimate heat sink by natural circulation, thus ensuring thecontainment integrity.
Internal Refuelling Water Storage Tank
The internal refuelling water storage tank is a passive water tank, resting on the internal
structure base slab. During refueling operations, it provides water for refueling cavity,internals storage compartment and refuelling transfer canal to complete the refueling
operation. Under the condition of loss-of-coolant accident and the steam pipe rupture, it
provides water for emergency reactor core cooling. In the severe accident, water in it floods
the internal structure under the balanced water level as a result of gravity. During theoperation of automatic depressurization system (ADS), it absorbs the sprayed steam from the(ADS). During the operation of the passive residual heat removal cooler, it works as the heatsink of the passive residual heat removal system.
Reactor pool
The reactor pool is used during refuelling operations or inspection of reactor internals. The
reactor pool has the following two compartments which can be separated by bulkhead:reactor cavity and internals storage pool adjacent to the reactor.
Containment system
The ACP100 containment houses the RCS, the passive safety systems and the auxiliary
systems. Small concrete containment with steel plate membrane encased and passive
containment heat removal system are adopted. The passive containment heat removal system
consists of two containment condensers, water storage tank and associated valves, piping andinstrumentation.
6. Plant safety and Operational Performances
Nuclear safety is always the first priority considered in design of ACP100 SMR that can besafely operated and provide clean energy. The ultimate goal of nuclear safety is to establish
and maintain an effective defence that can effectively protect people, community, andenvironment from radioactive disaster. To be specific, the design and operation of ACP100
ensures that radiation dose to the workers and to the members of the public do not exceed thedose limits and kept as low as reasonably achievable. Accident prevention measures ensure
25

that radioactive consequences are lower than limited dose in terms of all the considered
accident sequence and even in the unlikely severe accident, mitigation of accident induced
influences can be ensured by implementing emergency plan. The design of ACP100
incorporates operational experience of state of art design. Proven technology and equipment
are adopted as much as reasonably possible.
7. Instrumentation and Control systems
The Instrumentation and Control (I&C) system design for ACP100 will be based on defence
in depth concept, compliance with the single failure criterion and diversity. The diversity inthe design of I&C system is achieved through: (1) different hardware and software platforms
for 1E and N1E I&C, (2) the 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 processcontrol systems. The general control scheme of the NSSS is automatic control combined withmanual control. The control systems have an automatic control range from 0–100% fullpower, except for the reactor power control system and the SG feedwater control system. The
instrumentation system uses information from three (3) separate types (each type includes 4
independent channel) of instrument channel having source range, intermediate range andpower range to provide three separate protection and monitoring levels. The RPS has aconfiguration of four (4) redundant divisions. The functions of reactor trip and ESF actuation
are completed by the four (4) redundant protection divisions. Four (4) redundantmeasurements, using four (4) separate sensors, are made for each variable used for reactortrip.
8. Plant Arrangement
The two reactor modules share the fuel assembly building, the electrical building and nuclear
auxiliary building, as shown in Figure 3. Thus maximizing the sharing of facilities leads to a
reduced cost.
Reactor building
Reactor building and spent fuel pool of the ACP100 reactor are located under the groundlevel to enhance better protection from external events, reduce the radioactive material
release and provide better security.
Electrical building
The electrical building contains storage battery, power distribution equipment, I&C
equipment; main control room, remote shutdown station; ventilation system and smoke
exhaust system in the related rooms; chilled water system is employed for the ventilation of
the electrical equipment room.
Connection building
The connection building is meant to connect reactor building and other buildings, forconvenience of personnel passage, equipment transportation and containment dome
tensioning. There are pumps for residual heat removal system, pipes and components formain feedwater system and main steam generator system, as well as process systempenetration, I&C penetrations in the connection building.

26

Balance of plant:
Turbine Generator building
The turbine generator building 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 other side 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 andother auxiliary equipment.
Electric power systems
The turbine-generator (T/G) system is a double flow, tandem compound, with the lastrotating blades have the length of 800 mm. The most common speed of 3000 rpm is
adopted with an integrated reheat unit. The T/G train consists of a single flow
combined high pressure cylinder, one double-flow low-pressure cylinders, one
generator, one external MSRs with one stages of reheating are located on the left sideof the high pressure cylinder. Other related system components include a complete
T/G bearing lubrication oil system, a digital electro-hydraulic (DEH) control systemwith supervisory instrumentation, a turbine gland seal system, over speed protective
devices, turning gear, a turbine-generator stator coil, a turbine-generator seal oil
system, a rectifier section, and a voltage regulator.

Figure 3: The ACP100 plant layout diagram(Reproduced courtesy of CNNC)
9. Design and Licensing Status
The ACP100 engineering design is close to completion and a preliminary safety assessment
report (PSAR) was recently approved. Passive emergency cooling system, control rod drive27system, and critical heat flux have been tested. Control rod drive line cold tests and passive
containment cooling system tests are still underway. CNNC is to submit a project proposal to
the National Development and Reform Commission (NDRC) for approval. CNNC/NPIC has
built up comprehensive testing facilities which fulfil the needs of ACP100 design, and the test
results on crucial technology provided the necessary basis for final design and safety
evaluation of the reactor. Additionally, a nuclear power plant site safety assessment report anda siting stage environmental impact assessment report to China's regulator are supposed to besubmitted for review. In April 2015, an agreement to conduct a generic safety review for
ACP100 was signed between the IAEA and CNNC. An industrial demonstration plant withtwo 310 MW(th) unit is planned in Fujian Province, China.
10. Plant Economics
ACP100 is estimated to be competitive against large reactors if more than 4 modules are built
simultaneously. In feasibility study, the cost of ACP100 is around 5000 US$ per kW.

28

CAP150 (SNERDI/SNPTC, China)
1. Introduction
CAP150 is an integral type small modular reactor which employs the most advanced PWR
technology, developed by Shanghai Nuclear Engineering Research and Design Institute(SNERDI), a subsidiary of the State Nuclear Power Technology Corporation (SNPTC). It has
more simplified systems and more safety than current operating PWRs, to generate an electric
power of 150 MW(e). Figure 1 shows the CAP150 reactor system configuration.

Figure 1: Reactor System Configuration of CAP150(Reproduced courtesy of SNERDI/SNPTC)
2. Target Application
As a supplement to large reactors, the CAP150 is mainly developed for providing a flexible
way for remote electric supply and district heating, and also to replace the old thermal power
plants near cities.
3. Development Milestones
Nov 2009
May 2010
Dec 2013

Research work start
Conceptual design startedConceptual design finished29

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

SNERDI/SNPTCCountry of origin

China

Reactor type

Integral PWR

Electrical capacity (MW(e))

150

Thermal capacity (MW(th))

450

Expected Capacity Factor (%)

95

Design Life (years)

80

2Plant Footprint (m )10000

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)15.5oCore inlet/exit temperatures ( C)290/310Main Reactivity Control Mechanism

Control rod driving mechanisms and soluble boron

RPV Height (m)18

RPV Diameter (m)4.5

Module Weight (metric ton)<200 (Maximum component weight)

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Rankine cycle

Fuel Type/Assembly Array

UO2 pellet/15x15 square

Fuel Assembly Active Length (m)2.9

Number of Fuel Assemblies

69

Fuel Enrichment (%)

4.5 (average)

Fuel Burnup (GWd/ton)37 (average)

Fuel Cycle (months)

36

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

4

Refuelling Outage (days)<20

Distinguishing features

Integrated primary system. Passive safety.

Modules per Plant

2

Target Construction Duration (months)

36

Seismic design (g)0.30

Predicted core damage frequency per reactor year

< 1E-7

Design Status

Conceptual design finished

4. General Design Description
Design Philosophy
CAP150 design is based on the state of art philosophy of advanced PWR reactor concepts. It
aims to innovatively bring out a new G-III+ LWR with higher safety, reasonable andcompetitive economy, and good engineering feasibility.
The design principle includes:

30

•
•
•
•
•
•
•

Good engineering flexibility, multipurpose.
Employing proven technology and components to improve the reliability andeconomy.
Standardized and modularized Nuclear Steam Supply System (NSSS), the reactor
coolant system (RCS) modules are fully manufactured in factory and transported tosite for installation.
Defence in depth. Diverse safety systems include active and passive features. The
ultimate heat sink of passive long term cooling is air.
Enhance safety features of spent fuel storage.
Environment and public friendly, eliminating offsite emergency requirements.
The probabilistic safety analysis (PSA) and deterministic analysis are performed
simultaneously to achieve the safety goals from the early stage of design.

Nuclear Steam Supply System
CAP150 employs integral design, main equipment in the RCS including the pressurizer(PZR), steam generators (SGs), reactor coolant pumps (RCPs) and control rod drive
mechanism (CRDM) are all integrated into the reactor pressure vessel (RPV). Main pipelines
are eliminated. The RPV is 4.5 m in diameter and about 18 m in height.
Fuel and Reactor Core
Proven fuel assemblies of 300 MW(e)-class NPPs are used in this SMR. The FA300 fuelassembly is developed by SNERDI, and it has been widely used in Qinshan NPP(in China)
and Chashma NPP(in Pakistan), presenting very good mechanical and physical behaviour.
Figure 2 shows the schematic diagram of the fuel. The FA300 is suitable for a lower burnup
feature of SMR, it is currently available and good to shorten the development and reduce
R&D cost.

Figure 2: Fuel Assembly of CAP150(Reproduced courtesy of SNERDI/SNPTC)

31

There are totally 69 FAs in the core with 9x9 arrangements. The height and diameter of active
core is 2.9 m and 2m, respectively. Steel-water reflector is employed to improve the neutron
economy and reduce the irradiation damage to vessel.
Reactivity Control
There is no soluble boron in the core of CAP150. Instead, only control rods are used tocontrol the reactivity. The boron-free core eliminates systems for boron adjustment and PH
value control, simplifies the operation complexity and frequency of chemical control system,
greatly reduces the release of liquid radwaste, and avoids boron dilution accidents. The
design of boron-free core accords with requirements for small reactors like system
simplification, easy operation and low radioactivity. Another advantage of the mechanicalshim control is its quick transient responses and strong ability of load following. This makes
the SMR suitable for operating at variable loads, which is especially important for a small
size grid or a region with only single power supply. Electromagnetic CRDMs are installed on
the inner head of the RPV, and connected with control rod assemblies including adjustment
assemblies and safety assemblies.
Reactor Pressure Vessel and Internals
The RPV is assembled in three parts (upper, mid, lower), connected with flange joint. The
SG, PZR, RCP, CRDM are all installed in the mid-part. There are no penetrations on theupper and lower-part of RPV. The pressure boundary of the upper and lower RPV is simple
and safe. Because the RCP’s motor and CRDM are not directly in contact with the coolant,
these components will not suffer high temperature and high radiation. The RCP and CRDM
are based on current technology, and also possess higher reliability and maintainability. There
are double heads at the upper part of RPV. Two physical barriers ensure the integrity ofpressure boundary, and prevent coolant loss and radioactive release. Pressure is pre-charged
by high pressure nitrogen between the inner and outer head so as to reduce the probability
and consequence of rod ejection accidents. In addition, the pre-charged pressure can also
eliminate the probability of the any leakage and break of the many pipelines and penetrations
on the annular support platform and also the inner head. Figure 3 shows the schematic
diagram of integral pressure vessel and internals.
Reactor Coolant System
The primary coolant is heated and flows upwards in the core, goes through the riser, and thenfrom the upper parts enters the annular down-comer outside the barrel. Eight SGs are housedin the upper part of the down comer. The coolant is pumped downwards through the SGs andcooled down and then fed back into the core again.
Steam Generator
There are eight SGs, which are circularly distributed in the RPV down comer. The inlet of
feed water and outlet of steam are located on the annular support platform. The in service
inspection of the tubes to monitor failures or blocking are performed from the secondary side.
Each SG corresponds to one RCP. One of eight SGs failure will not cause significant impact
on normal operation. The design pressure of steam/feedwater pipelines is the same as that ofthe primary loop. Accidents such as steam generator tube rupture (SGTR) and steam line
break (SLB) can be quickly mitigated by closing SG isolation valves. In addition, SGs arealso used as heat exchangers of secondary side residual heat removal system.
Pressurizer
The pressurizer is located on the top of the riser. The CRDMs pass through the upper dome of32

the pressurizer. The saturated fluid in the pressurizer body is connected with the subcooled
coolant in the riser through surge holes setting in the bottom plate of the sleeve. Adjustmentof pressure in the pressurizer is achieved by heating and spraying, to control the pressure in
the primary loop.

Figure 3: Integrated RCS and components(Reproduced courtesy of SNERDI/SNPTC)
5. Safety Features
CAP150 has some inherent safety features, as shown in the following aspects:
•
•
•
•
•
•
•
•

RCS adopts integral design, which eliminates large pipelines and reduces small size
pipelines. This design basically eliminates large break LOCAs and reduces theoccurrence frequency of small break LOCAs.
Double heads of the RPV further increase the integrity of pressure boundary in the
primary loop and radioactivity barrier.
Separated configuration of multiple sets of RCPs and SGs decreases the impact of
SLB and rotor seizure when one set of these equipment failure.
The design pressure of steam/feedwater pipelines are the same as that of the primaryloop. Accidents such as SGTR and SLB can be quickly mitigated by closing isolation
valves.
Small flow resistances in the integral reactor, strong natural circulation ability, small
core power and low power density are helpful for the mitigation of accidents such as
station blackout and loss of flow accident.
Large water inventory and large thermal inertia ensure enough response time for
accident mitigation.
There is no soluble boron in the coolant, and thus the core has a better negative
feedback coefficient.
The steel containment is submerged in containment cooling pool. If some leakage
occurs, the radioactivity retention by containment cooling pool and concrete canfurther reduce the release of radioactivity to environment.

33

•The in-containment spent fuel pool assures continuous cooling of spent fuel after
accidents as well as containment of high radioactivity.

Engineered Safety System Approach and Configurationin addition to the inherent safety features, engineered safety systems are considered toenhance the safety. Both active and passive systems are employed to contribute a morediverse approach and decay heat removal paths. Figure 4 shows some passive safety systems
in the CAP150, including the passive decay heat removal by SG secondary side, passive
decay heat removal by primary side, simplified passive safety injection system, and passive
containment cooling system.

Figure 4: Passive safety system of CAP150(Reproduced courtesy of SNERDI/SNPTC)
Decay heat removal system
In normal operation, the reactor heat is removed by secondary loop. Feedwater system is nonsafety-related. It can be taken as the first heat removal channel to maintain the cooling of thecore. There are totally 8 SGs and only two of them are needed for coolant cooling after
accidents. The normal residual heat removal system can be put into use in both normal
operations and accident conditions. Two sets of the system are deployed. Each of them has
100% capability to remove core decay heat. Its pipelines and heat exchangers are designed toendure the primary pressure. Thus it can be used to remove the residual heat in emergency
conditions. The system is safety-related and can be taken as the secondary heat removal
channel to cool the reactor. Both passive residual heat removal systems in the primary loop
and that in the secondary side are safety-related systems. These two systems may be initiated34

at the same time by safety signals. They may be taken as the third heat removal channel.
Emergency core cooling system
The safety injection system of the CAP150 is greatly simplified compared with operational
PWRs. In current PWR power plant, the safety injection system has three main functions.
The first is to provide emergent and durative cooling for the core when LOCA occur. The
second is to compensate the reactor coolant inventory for small leaks or depressurizations tomaintain water level in pressurizer while chemical and volume control system (CVCS) is
failed. The third is to inject high concentrated boron solution into the core so as to prevent thecore from re-criticality when over cooling events such as SLB takes place. In order to achievethese three functions, current safety injection system has three subsystems: high pressure
injection, accumulator and low pressure injection. The high pressure injection implements thesecond function while CVCS is invalid and the third function when SLB happens. When
LOCA takes place, the flow offered by high pressure injection is too low to maintain core
covered. Accumulator is then needed to provide a very high flow rate injection at a middle
pressure level. As the system depressurizes to a pressure low enough, the low pressure
injection starts and remove decay heat for a comparative long term. The design features of
CAP150 make it possible to greatly simplify its safety injection system. The design pressureof the normal residual heat removal system is the same as that of the primary loop. The
second function of the safety injection system can be performed by the normal residual heatremoval system. Since the seriousness of the SLB accident decreases, the optimized design ofcontrol rod assembly makes the core shutdown margin enough and core retune to criticalimpossible. Thus the third function of the safety injection becomes unnecessary. The only
requirement left for safety injection system of CAP150 is to mitigate small break LOCA. For
CAP150, large break LOCA is eliminated because of integrated RCS so that accumulator
injection is not quite necessary. The core power is limited and the power density is quite low.
The specific coolant inventory is very large compared to current PWRs. The containment freevolume is small enough that the pressure between reactor vessel and containment can reachbalance very soon after small break LOCA. Even if there is no high pressure injection, thecore will not uncover before low pressure injection starts. Hence for CAP150, it is believedthat only low pressure injection is needed to maintain core safety. Therefore, only the lowpressure injection system is deployed in CAP150. As shown in Figure 4, it consists of arefuelling water tank, auxiliary depressurization system and the related valves and pipelines.
Containment system
The upper part of the containment is submerged in the containment cooling water pool. When
LOCA or SLB accident occurs, steam blows into the containment, steam condenses on thewall of the steel containment, and the heat is transferred to the containment cooling pool. The
condensed water partly returns to the refuelling tank for the low pressure safety injection, andpartly flows into the sump. The water in the containment cooling pool will be evaporated andthe water inventory can last 7 days without any human intervention. In addition, there is a
supplement pool to make up for the containment cooling pool when its water level becomes
low so as to keep the passive cooling of the containment for a long term. This containment
cooling system doesn’t need actuation signals and can perform its safety functions as soon as
accidents occur. It runs fully by natural circulation. The principle of the system is simple andreliable. Moreover, the containment cooling pool provides an additional barrier forradioactivity and thus further decreases release possibility. The spent fuel storage pool islocated in the containment, so that the safety of spent fuel under SBO is ensured by thepassive containment cooling systems. And all radioactivity is contained in the containment
during normal operation or an accident.

35

6.

Plant safety and Operational Performances

Using small event tree / large fault tree methodology, preliminary level one PSA analysis ofinternal events at power for CAP150 was carried out. Based on analyses and postulations on
initiating events, event tree models and success criteria, corresponding model was built
including initiating events frequency, event tree models, fault tree inputs, human error andcorresponding processing and so on. Preliminary quantified analysis results show that thecore damage frequency (CDF) is about 3.5x10-8/reactor-year. LOCA events are the largest
contribution accounting for about 40%. It is estimated that the large release frequency (LRF)
of CAP150 is about 10-9/reactor-year in magnitude or even less because of several design
features such as the retention of core debris and passive containment cooling are considered,
the core power level is low, the time without human intervention is relatively long, the wholecontainment is submerged in the containment pool so that one more barrier against the large
radioactivity release is added etc.
7. Design and Licensing Status
Conceptual design completed.

36

CAP200 (SNERDI/SNPTC, China)
1. Introduction
The China Advanced Passive pressurized water reactor 200MW(e) (CAP200) is one of theserial research and development products of PWRs adopting passive engineered safety
features initiated by SNERDI. The design of CAP200 is based on the experience of the PWR
technology R & D for more than 45 years, construction and safe operation for more than 20
years in China. It is the outcome of accumulated experience and achievements of the world's
first batch of AP1000 units and the R&D of CAP1400. Furthermore, it adopts safetyenhancement measures based on lessons learnt from Japanese Fukushima nuclear accident.
Figure 1 shows reactor system configuration of CAP200.

Figure 1: Reactor System Configuration of CAP200(Reproduced courtesy of SNERDI/SNPTC)
2. Target Application
CAP200 can be used as a supplement to large PWRs, this reactor is designed for multiple
applications, such as nuclear cogeneration and replacing retired fossil power plants in urban
areas.
3. Development Milestones
May 2014

Conceptual design started

December 2015

Conceptual design finished

4. General Design Description
Design Philosophy
CAP200 is a small PWR which is designed with improved safety, flexibility andenvironmental friendliness, and is also comparable with other SMRs on economy. Compared37

with large PWRs, CAP200 has a number of advantages such as higher inherent safety, lower
frequency of large radioactivity release, longer time without 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:• Compact layout primary system: Steam generators (SG) are connected to reactor
pressure vessel (RPV) directly and main pipes are eliminated. Compact layout ofsystem and components results in lower risk and possibility of loss of coolant
accidents and smaller primary system footprint.
• Modularized: main modules can be fabricated in factory and transported to the site for
installation. Construction period can be shortened because of high modularization andthe use of existing infrastructure.
• Redundant and diversified passive safety features: redundant and diversified active
and passive safety features are deployed, which ensures the reactor core safety andextremely low risk of large radioactivity release.
• Steel containment is located below ground level, which eliminates the risk of outside
attacks and improves the ability of preventing natural disasters.
Primary circuit and its main characteristics
CAP200 has two primary coolant loops which transfer heat generated by fission reaction of
U235 in reactor core to SG. For each loop, it consists of one SG, one main coolant pump andpressure nozzles connecting them. A pressurizer (PZR) is connected to reactor coolant system
by surge line. The SG is located at higher elevation than RPV in order to achieve natural
circulation. The elevation of PZR is higher than that of pressure nozzles which enhances thedrainage of PZR liquid to the reactor coolant system and reduces the carry-over of liquid to
PZR when the depressurization system is activated. The design of surge line minimizes thepotential for thermal stratification. As the barrier against release of radioactivity, the reactor
coolant system pressure boundary is designed to provide a high degree of integrity for
operation of the plant. CAP200 employs compact design, which means that main equipmentin the reactor coolant system (RCS) including the RPV, SGs and reactor coolant pumps(RCP) are connected by short pressure nozzles so that main pipelines are eliminated. RPVand SG are connected by a straight nozzle and SG and RCP are connected by an elbow. In
each pressure nozzle an inner duct is cased. Hot water leaving reactor core flows through theinner circular duct into hot channel of SG. After heat release in the SG heat transfer tubes,
cold water flows from cold channel of SG into pump suction through inner elbow duct and ispumped through the outer annular elbow back into third chamber of SG and then into
downcomer of RPV through outer annular duct.
Reactor coolant pump
Both leak-tight canned motor pump or wet coil pump are possible choices for the RCP of
CAP200. There is rich 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 65m at design flowrate of 12000 m3/h with a cold leg temperature of 289oC. The reactor coolant pump has no
shaft seals, eliminating the potential for seal failure LOCA, which significantly enhances
safety and reduces pump maintenance. The pumps have an internal flywheel to increase thepump rotating inertia and thereby providing a slower rate-of-flow coastdown to improve core
thermal margins following the loss of electric power.

38

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

SNERDI/SNPTCCountry of origin

China

Reactor type

PWR

Electrical capacity (MW(e))>200Thermal capacity (MW(th))

660

Expected Capacity Factor (%)>93

Design Life (years)

60

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)15.5Core inlet/exit temperatures (oC)289/313Main Reactivity Control Mechanism

Control rod drive mechanisms and soluble boron

RPV Height (m)8.545

RPV inner Diameter (m)3.280

Module Weight (metric ton)<200 (Maximum component weight)

Configuration of Reactor Coolant System

Compact

Power Conversion Process

Rankine cycle

Fuel Type/Assembly Array

UO2 pellet/17x17 square

Fuel Assembly Active Length (m)2.4

Number of Fuel Assemblies

89

Fuel Enrichment (%)

4.2 (average)

Fuel Burnup (GWd/ton)37 (average)

Fuel Cycle (months)

24

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

2

Refuelling Outage (days)<20

Distinguishing features

Compact layout; Passive safety; Underground containment.

Modules per Plant

1

Target Construction Duration (months)

36

Seismic design (g)0.3

Predicted core damage frequency per reactor year

<1E-6

Design Status

Conceptual design finished

Reactor core and fuel design
The reactor core and fuel serves as the source of heat generation in the reactor coolant
system. The reactor core consists of fuel assemblies, rod cluster control assemblies, grey rod
cluster assemblies, burnable absorbers, neutron sources, and thimble plugs. The fuelassembly is typical 17×17 lattice type used in commercial PWR, with shorter length and partof the grids number reduced. The assembly structure consists of cladding material, and UO2
ceramic pellet, with an enrichment of less than 5%, complying with regulation rules for civil
nuclear facility. IFBA burnable poison is present in the assembly to suppress the initialexcess reactivity. The core rated thermal power is 660 MW(th) with 89 fuel assemblies. The
39

core has an equivalent diameter of 2.29m, and the total uranium load is about 26t. Highburnup and low neutron leakage reactor core fuel management strategy has been used for
CAP200. For the advanced first core design，there are three radial regions that have different
enrichments ranging from 2.00 to 4.00% U235 to enhance the fuel economics. The active fuellength of reactor core is 240 cm. The linear averaged power density and volume averaged
power density for CAP200 are 11.4kW/m and 66.9MW/m3, respectively. The lower power
density leads to lower fuel center temperature, both in normal operation and transients. As aresult, the core has a good intrinsic safety feature with regard to the fuel. The core is designed
for a fuel cycle of 24 months with 93% availability, and averaged discharge burnup of
assembly is as high as 42218MWd/tU for balance cycle. CAP200 has at least 15% margin for
departure from nucleate boiling (DNB). The power reactivity coefficients, including fuelDoppler, moderator temperature and moderator void coefficients are negative providing
negative reactivity feedback characteristics and meeting the design criteria, which enhances
safety of reactor core inherently. CAP200 have enough margins for several operation modes
with the maximum value control rod withdrawal. The peak power factor of CAP200 satisfies
the limiting value for every fuel cycle. Core reactivity is controlled by both soluble boron andcontrol rods. CAP200 is capable of load following without boron dilution. Nevertheless, thecontrol 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 theconventional boron system and the dilution operation. The CAP200 reactor core adopts two
different types of control rod assemblies. There are 37 assemblies in total. The high worth
assemblies are known as “black” rods. These are used for shutdown and large swings ofreactivity. The grey rods with lower worth are used for load-follow during operation of a fuelcycle to avoid adjusting soluble boron concentration, which results in a substantial reduction
in waste water generation and treatment for PWR required to execute load-follow operation.
Reactor pressure vessel
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 RPV is a high-pressure
boundary used to support and enclose the reactor core. The vessel is cylindrical with aremovable flanged hemispherical upper head and hemispherical bottom head. The height of
RPV is 8.545m and inner diameter of reactor core region is 3.280m. The area with highest
neutron fluence of active core region of the vessel is completely 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 the reactor
through upper head of reactor vessel which eliminates penetrations in the lower head of RPV.
This reduces the possibility of a loss of coolant accident (LOCA) and uncovering of the core
The core is positioned as low as possible in the vessel to decrease re-flood time in the eventof an accident; furthermore, this arrangement is very helpful for successful execution of IVR.
Reactor internals
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 ofstainless steel is welded which eliminates the occurrence of reactor core damage induced byloosening of baffle bolts. The reactor internals design of CAP200 takes the inner pressure
nozzle installation into consideration.
Integrated head package
The integrated head package (IHP) is installed on the top of the reactor vessel head, and40

combines several components into one assembly to simplify refuelling operation anddecrease the radiation dose for refuelling and maintenance personnel. The IHP design of
CAP1400 renders a reference for that of CAP200. The CRDM cooling system, including
ventilation panels, fans and its supports, usually used in the conventional PWRs are removed
for CAP200 since the high temperature performance of coil in CRDM of CAP200 can
withstand the temperature of 350℃. The structure of IHP of CAP200 is much simpler. It
includes shroud assembly, lift rig assembly, radial arm hoist assembly, CRDM seismic
support, cable and cable support structure.
Steam generators
The steam generator (SG) of CAP200 is vertical U-tube type with 3861 tubes made ofthermally treated nickel-chromium-iron Alloy 690 on a triangular pitch, to fulfil therequirement of heat transfer capacity. The design parameter of steam exit from SG is 6.02
MPa and 183.2 kg/s for thermal design flow rate with no tube plugged, respectively. Pressure
nozzles are used to connect RPV and RCP directly with SG. The inner duct of the pressure
nozzle connecting RPV with SG side is welded to the partition board of SG water chamber.
The nozzles are designed with two ducts. For the nozzles between RPV and SGs, hot waterflows in the inner circular duct and cold water in the outer annular duct. Stainless steel bars
have been installed in 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 andadopts the separated start-up feedwater pipe which eliminates thermal stratification andprevents occurrence of water hammer. The SG channel head is divided into three parts, hotchannel, cold channel, and a third channel through which the coolant is pumped back into
reactor vessel.
Pressurizer
The pressurizer of CAP200 is a typical steam-type one with electrical heater at the bottom
and spray at the top. The pressure control is steady and reliable. The pressurizer is a verticallymounted cylindrical pressure vessel with hemispherical top and bottom heads which adopts
the traditional design that is based on proven technology. The pressurizer volume is designed
to be large which increases margins for transient operation and minimizes the number of
unplanned reactor trips and provides higher reliability. In addition, 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 responseby using large volume pressurizer.
Control Rod Drive Mechanism
The newly developed control rod drive mechanism (CRDM) of CAP200 will be able tooperate the motor and coil assembly at higher temperature. The heat generation and heattransfer coefficient of the CRDM coils are improved to reduce the cooling air flow rate,
which will also contribute to the simplification of the IHP since the cooling fan and related
seismic loads can be reduced.
5.

Description of safety concept

Safety requirements and design philosophy
The fundamental design philosophy of CAP200 is defence in depth including accident
resistance, core damage prevention and mitigation. Accident prevention is realized by using
the best available material and water chemistry, high quality requirement for components
during design and construction, negative reactivity coefficients, improved man-machineinterface system and extended operator response time. In addition, the need for offsite

41

emergency response shall be eliminated. For core damage prevention, it is mainly achievedby passive safety features which are adequate to automatically establish and maintain corecooling and containment integrity for a significant period of time after initiation of designbasis events assuming the most limiting single failure, no operator action, and with no onsite
and offsite ac electrical power sources. In order to mitigate accident, containment systems areprovided for heat removal and retention of fission products. Hydrogen control system isprovided to achieve that the hydrogen concentration in containment does not exceed 10%
under dry conditions for an amount of hydrogen equivalent to that generated by oxidation of
100% of the active fuel clad. In addition, redundancy has been used for containment
components, such as containment isolation valve. For CAP200, defence in depth is also
realized by some non-safety systems which decrease the events that may lead to core damage.
The core damage frequency and large release frequency are aimed at lower than 1.00x10-6
and 1.00x10-7, respectively.
Safety systems and features
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 forplant in simplicity, safety, O&M, availability and investment protection. No active
components such as pumps, fans and other machinery are used. A few simple valves alignand automatically actuate the passive safety systems. The passive safety systems are designed
to meet criteria of single failure, independence, diversity, multiplicity. For CAP200, passive
core cooling system and passive containment cooling system are adopted.
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 waterpool. The water pool is safety-related and prevents the containment from exceeding thedesign pressure and temperature following a postulated design basis accident by cooling theoutside surface of containment, as shown in Figure 2. The inventory in the water pool canlast at least 7 days after an accident. The Passive containment cooling system works without
operator control or external assistance. The passive containment cooling system performs thefollowing safety-related functions:
•
•
•
•

•

42

Reduce the containment temperature and pressure following a LOCA or main steamline break (MSLB) accident inside the containment by removing thermal energy fromthe containment atmosphere.
Retain fission product by gravitational deposition, thermophoresis and condensation
to wall of containment steel, all of the working mechanism are passive.
Transfer heat to the safety-related ultimate heat sink for other events resulting in asignificant increase in containment pressure and temperature.
Limit releases of radioactivity (post-accident) by reducing the pressure differential
between the containment atmosphere and the external environment, thereby
diminishing the driving force for leakage of fission products from containment toatmosphere.
Provide a seismic source of makeup water to the spent fuel pool in the event of a
prolonged loss of normal spent fuel pool cooling and provides a seismic source of
makeup water to the fire protection system.

Passive core cooling system
The main function of passive core cooling system is to provide emergency core cooling
during postulated design basis accidents by supplement and boration to RCS after non-LOCA
accidents and safety injection to core 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 isreferred as the Passive Core Cooling System. The Passive Safety Injection System consists of
two core makeup tanks，one in containment refueling water storage tank (IRWST), IRWST
screen, containment recirculation screen, pH adjustment basket and associated valves, piping,
instrumentation etc., as shown in Figure 2. The passive decay heat removal system consistsof one passive residual heat removal heat exchanger and associated valves, piping, andinstrumentation The heat exchanger is located in the IRWST, which works as the heat sink
for the heat exchanger.

Figure 2: Passive Core Cooling System and Passive Containment Cooling System(Reproduced courtesy of SNERDI/SNPTC)
Long-term accident mitigation
The passive containment cooling system will remove the heat from containment within 7days after the initiation of a postulated accident and keep the temperature and pressure
without exceeding the design values assuming no operator action and on-site or off-site ac
power. Within 7days, the inventory from the water pool guarantees that the containmentpressure won’t exceed the design value. After that the decay heat is low enough for the heat
pipes to transfer decay heat from the water pool to atmosphere and keep the water pool levelnot decreasing. In addition, supplement to the water pool can be obtained from a coolingwater storage tank. The non-safety related cooling water storage tank and the heat pipes keep
the containment cooled for indefinite duration.

43

Hydrogen control system
During degraded core accident, hydrogen will be generated at a greater rate than during thedesign basis LOCA, and the hydrogen control system is designed to deal with the risk. The
hydrogen control system consists of hydrogen monitoring system and Passive Auto-catalytic
Recombiners (PARs) system. The hydrogen detectors are installed in the region ofcontainment dome to measure the concentration of hydrogen which may be transferred tomain control room and remote shutdown station. Passive auto-catalytic recombiner system isinstalled to control the hydrogen concentration in containment without exceeding regulatory
requirement.
In-vessel retention of molten core debris
The IVR of molten core debris is one of the key mitigation strategies during severe accidents
for CAP200. During postulated severe accidents, flooding the reactor cavity with theinjection of IRWST water to submerge the RPV will be performed to implement the IVR
strategy and to maintain the integrity of RPV by cooling the external surface of the vessel.
This strategy prevents molten debris from relocating to the containment which may cause
further ex-vessel severe accident phenomena such as ex-vessel steam explosion and molten
core-concrete interaction.
Lessons learned from Fukushima nuclear accident
The resistance capacity of CAP200 to extreme natural hazards such as earthquake, flood andetc. has been enhanced. Since the containment is buried underground, CAP200 can also
withstand the impact of aircraft crash. Especially, without being dependent on alternating
current, the passive safety system is able to keep CAP200 safe within 7 days after accident
initiation; After 7 days, the non-safety grade measures are available to offer reactor corecooling for indefinite duration; In addition, the reactor core can still be cooled with someextra off-site assistance.
6. Nuclear island layout
The size of Nuclear Island is minimized by both system simplification and adoption ofpassive design. The number of system component is reduced by adoption of large capacity
equipment, common use of single equipment for different systems.
Maintenance
Building

Aux.
Building

Reactor
Building

Auxiliary
Building

Figure 3: Layout of Nuclear Island (Reproduced courtesy of SNERDI/SNPTC)
7. Design and Licensing Status
Conceptual design finished.

44

AHWR-300 (BARC, India)
1. Introduction
The Indian Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type lightwater cooled and heavy water moderated reactor to generate 300 MW(e). The reactor concept
adopts the proven pressurized heavy water reactor (PHWR) specific technologies pertaining
to pressure tube and low pressure moderator based design. The reactor is designed anddeveloped by Bhabha Atomic Research Centre (BARC). The AHWR aims to produce most of
its power from thorium, utilizing large-scale thorium for future commercial nuclear power
plants. The shop assembled coolant channels of the AHWR enables quick replacement ofpressure tube without affecting other installed coolant channel components. The layout
diagram of the AHWR plant is depicted in Figure 1. The reactor incorporates several
advanced passive safety features. The AHWR is designed to provide protection against
external as well as internal security threats, including insider malevolent acts. The plantdesign contains the following features:
•
•
•
•
•
•
•
•
•
•

Heat removal from core under all operating conditions including shutdown condition
by natural circulation of coolant;
Passive containment cooling (PCC) and containment isolation during loss of coolant
accident (LOCA);
Passive removal of decay heat of reactor in case of station blackout by isolation
condensers immersed in a large pool of water in a gravity driven water pool (GDWP)
without requiring any external source of power to open valves or pumps;
100-year target design plant life;
Direct injection of cooling water by emergency core cooling system (ECCS) in a
passive mode directly inside fuel cluster;
Passive poison injection into the moderator in the event of non-availability of both the
shutdown systems due to wired system failure or malevolent action;
Passive cooling of moderator and endshield in case of station blackout;
Dedicated core catcher to cool the core melt in case of very low probable severe
accident;
Hardened filtered vent system to depressurize the containment and retain radioactive
nuclides in case of severe accident;
A design objective requiring no exclusion zone beyond plant boundary on account of
its advanced safety features;

2. Target Applications
The AHWR is designed for electricity generation as well as sea water desalination, with atarget to produce 2400 m3/day of demineralised water. It can also be configured to deliver
higher desalination capacities with some reduction in electricity generation.
3. Development Milestones
2003 September

Peer review of the design performed

2006

Design review completed

45

GRAVITY DRIVEN WATER POOLSTEAM DRUM

FUELLING MACHINECALANDRIA

INCLINED FUEL TRANSFER MACHINEREACTOR BUILDINGFUEL BUILDING

Figure 1: Layout diagram of an AHWR plant (Reproduced courtesy of BARC)
4. General design description
Design philosophy
The AHWR is designed and developed for large-scale use of thorium in commercial nuclear
power generation with no external input of U233 in the equilibrium cycle. The AHWR
incorporates inherent and passive safety features to the maximum extent as a part of the
defence in depth strategy. The AHWR design aims to establish a case for elimination of theneed for evacuation planning in case of accident scenario. The AHWR passive and active
hybrid safety systems designed to cope with the consequences of design basis accidents. The
AHWR is also designed for seawater desalination. The AHWR schematic diagram is depicted
in Figure 2.
Nuclear steam supply system
The nuclear steam supply system (NSSS) consists of the reactor core and internals, coolant
channels, tail pipes and steam drums. The main heat transport (MHT) system transports heatfrom fuel pins to the steam drums, using light water as the coolant under natural circulation.
The MHT system consists of a common circular inlet header from which feeders branch out
to the coolant channels in the core. The outlets from the coolant channels are connected to
tailpipes carrying steam-water mixture from the individual coolant channels to the four (4)steam drums.

46

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

Bhabha Atomic Research Centre (BARC)Country of origin

India

Reactor typePressure tube type heavy water moderated reactor(vertical flow)

Electrical capacity (MW(e))

304

Thermal capacity (MW(th))

920

Expected Capacity Factor (%)

90

Design Life (years)

100

Plant Footprint (m2)

220000

Coolant/moderator

Light water/heavy water (D2O)

Primary circulation

Natural circulation

System pressure (MPa)7.0oCore inlet/exit temperatures ( C)

258.2 / 285 with average steam exit quality of 19.7%

Main Reactivity Control Mechanism

Rod insertion

Calandria Height (m)5Calandria Diameter (m)6.9Calandria Weight (ton)50

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Rankine Cycle

Fuel Type/Assembly Array

Fuel Assembly Active Length (m):Flexibility in fuel options: Can use Either (Pu-Th)MOX and (ThU233)MOX or ((Th-LEU) MOX with LEU of 19.75% U235
enrichment/54 pins (12, 18, 24)
3.5

Number of Fuel Assemblies:452

Fuel Enrichment (%):

4.21% average fissile content

Fuel Burnup (GWd/ton):

60

Fuel Cycle:

Th-Pu-U233 or LEU - Thorium

Cogeneration Capability:

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains:

4

Refueling Outage (days):

On-power

Distinguishing features:

Natural circulation driven heat removal during normal operation andhot shutdown, Passive containment cooling and isolation system
during LOCA, Passive corium cooling by core catcher, Double
containment system. Passive moderator and end shield cooling
during SBO. Passive decay heat removal using ICs in case of SBO

Modules per Plant:

1Target Construction Duration(months):~60

Seismic design (g):0.20

Predicted core damage frequency (per reactor year):~ 1 E-8

Design Status:

Basic design completed

Reactor core
The reactor core is housed in calandria, a cylindrical stainless steel vessel containing heavy
water which acts as a moderator and reflector. The core consists of a total 513 lattice47

locations arranged in square pitch of 225 mm. There are452 coolant channel assemblies in thecore, and it encloses the pressure tubes that are arranged in a vertical orientation to facilitate
natural circulation flow. The circular fuel cluster of the AHWR shown in Figure 3 contains 30(Th, U233 MOX pins and 24 (Th, Pu) MOX pins, along with a displacer rod at the centre.
REACTOR BUILDINGGRAVITY DRIVEN
WATER POOL(GDWP)PASSIVE CONTAINMENTCOOLING SYSTEMISOLATION CONDENSERPASSIVE CORE DECAY HEATREMOVAL SYSTEMSTEAMSTEAM DRUM
(4 Nos.)H2, NMCA AND Zn INJ

TURBINE

GENERATORACCUMULATOREND SHIELDCOOLINGSYSTEMGDWP
INJECTIONDESALINATION
PLANT

FEED PUMP
DEAERATORECC HEADER

FEED
WATER
HEATERSCOOLING
WATER

CEP
INLET
HEADER

GDWP
COOLINGSYSTEM

CONDENSERTURBINE BUILDINGCALANDRIA VAULT
COOLING SYSTEM

PROCESS WATER

Figure 2: Schematic diagram of AHWR (Reproduced courtesy of BARC)
Reactivity control
The AHWR reactivity control is performed through the use of twenty four (24) numbers ofcontrol rods, i.e., eight (8) absorber rod, eight (8) regulating rods and eight (8) shim rods
respectively. For the reactor control and shutdown, tantalum alloy based control and shut-off
rods are provided. The system consists of a set of shut-off rods, and held on top of the reactor
core by individual electro-magnets. These shut-off rods are passively released underabnormal conditions. This is a fail-safe system, so that in case of loss of power from batteries,
the shut-off rods would fall under gravity and shutdown the reactor.
Reactor coolant system
The AHWR primary cooling mechanism under normal operating condition and shutdown
condition is by natural circulation of coolant. The reactor coolant system (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 calandria, located below ground levelcontains vertical coolant channels in which the coolant picks up heat from the fuel assemblies
suspended inside the pressure tubes. The naturally circulated coolant flow through the tailpipes to the steam drums, where steam is separated for running the turbine cycle. The four (4)steam drums, receive feed water at stipulated temperature to provide optimum sub-cooling atreactor inlet. Four (4) down-comers from each steam drums bring the flow to a circular inlet
header which distributes the flow to each of the 452 coolant channels through individual
feeders.

48

Figure 3: AHWR reactor block and fuel cluster (Reproduced courtesy of BARC)
Steam drum
Steam drum is a horizontally mounted cylindrical vessel closed at both ends by the tori
spherical heads. The two-phase steam-water mixture produced in the reactor core enters thesteam drums through tail pipes connected to the coolant channels. The feed water enters thesteam drum through a sparger, which runs along the length of the steam drum and located in
the space between the partition plates. Steam is taken out from each steam drum throughoutlet nozzles located on the top of the steam drum.
5. Safety Features
Engineered safety system approach and configuration
Several advanced safety features are adopted in the AHWR design to enhance the safety andeliminate any potential impact in the public domain in the case of safety events. The AHWR
design also incorporates several passive systems for performing safety-related functions in
the event of an accident.
Decay heat removal system
The shutdown cooling system is capable to remove the decay heat and cool down the mainheat transport system (MHTS).The main steam condenser (MSC) removes the decay heat andcool down the MHTS. Normal cool down from 558 K to 423 K is achieved by MSC or
passively by isolation condensers (ICs), when the MSC is not available. During hot shut
down condition, steam generated due to decay heat is fed to the MSC. This steam after
condensation is pumped back to the steam drums by feed water pumps. The ICs submerged in
the GDWP dissipates the heat to the GDWP.

49

Emergency core cooling system
The emergency core cooling system (ECCS) consists of four (4) independent circuits and isactuated in passive mode as depressurisation of the MHT system progresses following a
LOCA. ECCS operation consists of high pressure injection system using accumulators and alow pressure injection system using GDWP as source of water in a sequential manner. The
AHWR design provides a grace period of 7 days for absence of any operator or powered
actions in the event of accident.
Gravity driven water pool
The gravity driven water pool (GDWP), a part of the ECCS is a large pool of watercontaining 8000 m3 of water inventory, located in the dome region of the reactor building toaccommodate the cooling requirements of different systems during various reactor
conditions. The GDWP consists of eight (8) interconnected compartments. The GDWP is
provided for:• Removal of core decay heat during shutdown (hot and cold shutdown) bycondensation of steam flowing through ICs that are submerged in GDWP;• Direct injection of low pressure coolant into the core after LOCA for 3 days;• Vapour suppression during LOCA;• Removal of containment heat by condensation of steam from the air-steam mixture
across the PCC tubes during and after LOCA.
• Removal of heat from moderator and endshield in case of SBO.
The GDWP recirculation and cooling system is provided for cooling the GDWP inventory
catering to the above requirements as well as recirculating, filling and draining the water ineach compartment of the GDWP. It consists of four (4) heat exchangers, four (4) pumps,
filters, ion exchangers and a chemical addition tank to maintain the water chemistry.
Reactor protection system
The reactor protection system comprise of two (2) independent fast acting shutdown systems(SDS).SDS-1 is based on mechanical shut-off rods with boron carbide based absorbers in 33
lattice positions, providing a total negative reactivity worth of 74 mk with all rods inserted,and a worth of 51 mk with maximum two (2) worth rods not available. SDS-2 is based on aliquid poison injection into the moderator. In addition, a pressurized addition of poison
passively driven by steam pressure takes place in the event of over pressure in the MHT
system. For a long-term sub-criticality control, there is a provision to add boron to themoderator.
Containment system
The containment system houses the reactor, the primary coolant and moderator systems andother systems connected with steam generation. The containment system is designed with
passive and inherent safety systems. The primary containment is designed with a diameter of
52 m, height of 75.5 m (54.5 m above ground) and a design pressure/temperature of 0.274
MPa/410 K. Double containment cylindrical concrete structures, roofed with two segmental
concrete domes are provided. The GDWP inventory is provided for removal of containment
heat by condensation of steam from the air-steam mixture across the PCC tubes, during andafter LOCA. Following a large break LOCA the double containment, passive containment
isolation system and passive containment cooling system help in passively bringing down thecontainment pressure and in minimizing any releases from the containment. For the purposeof containment design, a double-ended guillotine rupture of the 600 mm diameter inlet header
has been considered as the design basis accident. The hydrogen mitigation system and hardvent for venting of containment and retention of fission products is incorporated in the50

primary containment in response to the Fukushima Daiichi event.
6. Plant safety and operational performances
The AHWR is designed to operate efficiently at all operating conditions using natural
circulation as the means of providing core coolant flow. The natural circulation flow ratedepends on the operating conditions such as the operating pressure, power and feed waterflow rate and its temperature. The natural circulation flow rate has also been predicted atpartial power conditions considering the feed water temperature variation and its effect. A
separate cold start-up procedure for the reactor is required to avoid low power flow
oscillations. As per the start-up scheme, boiling in the MHTS is permitted only at 7.0 MPa.
The desired pressure in the steam drum is maintained by supplying steam from the start-up
boiler. Under normal operating conditions, the pressure in the steam drum is maintained at
7.0 MPa. The feed water with the same mass flow rate enters the steam drum at 403 K. The
primary circulation flow rate is of the order of 2145 kg/s maintaining the average steam exit
quality at about 19%.The target lifetime load factor and availability factors for the AHWR are
80%and 90%, respectively.
7. Instrumentation and controls systems
The instrumentation and control (I&C) of AHWR uses modern electronic technology and all
the key real time control and monitoring systems are designed around computer based
platforms. The general concepts and design principles followed are redundancy, diversity, failsafe behaviour, fault tolerance, testability and maintainability. The I&C system is designed toregulate the reactor power and process variables of the plant within limits under all operating
conditions and postulated emergency situations. The 452 channels in the AHWR are
instrumented with in-house developed bi-directional venture flow meters for monitoring theforward flow and identify flow reversal. The design objectives met by the philosophy ofdefence-in-depth resulted in high reliability and availability of I&C systems, fulfilling
stringent safety and operational requirements.
8. Plant Arrangement
Reactor building
The reactor building is a double containment cylindrical concrete structure roofed by twosegmental concrete domes. The double containment structure besides containing andcontrolling release of effluents also provides necessary radiation shielding to thesurroundings from the radiation sources present in the reactor building during normal
operation. The GDWP is located near the top of the reactor building. The reactor building air
supply and exhaust ducts are shaped in the form of U-bends of sufficient height for passive
containment isolation.
Control building and structures
The AHWR service building provides nuclear and conventional service facilities likeworkshops, ventilation intake, compressors and decontamination areas as well as laboratory
services of chemical and instrumentation to power station complex. The main control
building housing safety group-1 I&C hardware is functionally and physically independent
from the backup control building housing safety group-2 I&C hardware.

51

Balance of plant:
Turbine Generator building
Turbine building accommodates turbine generator, condenser and associated steam
cycle equipment required for generation of electricity from steam cycle of the plant.
At normal full load, a flow of 408 kg/s, 0.25% wet saturated steam, at a pressure of
6.83 MPa from steam drum is delivered to the turbine throttle. After expansion in thehigh pressure turbine, steam is exhausted at pressure of 1.025 MPa. Then it passes
through an external moisture separator reheater. Subsequently the steam is reheated inbled steam reheaters and enters the low pressure double flow turbine at temperature
503.2K and pressure of 0.956 MPa, where it expands to a condenser back pressure of
8.3kPa.
Electric power systems
AHWR electrical power supply system comprises of off-site supplies and on-site
supplies, and associated distribution systems. The onsite power supplies are divided
into normal and emergency power supply systems. The emergency power supply
systems are only considered as safety related systems. The various unit auxiliary loads
connected to the electrical power supply system are broadly divided into two
categories: safety and non-safety related loads. Non-safety related loads can endure
prolong power failure without affecting the safe operation of plant. The safety relatedloads depending on their criticality have defined limitation of duration of power
failure to ensure safe operation of plant. The emergency power supply system ensureselectrical power with specified quality to safety related loads which can take care ofsafe operation or shut down of plant after a postulated initiating event.
9. Design and Licensing status
The AHWR pre-licensing review has been completed by the national regulatory body which
has approved AHWR design in principle. The regulatory body has recommended forexperimental demonstration of the first-of-a-kind features. The Nuclear Power Corporation ofIndia Ltd. (NPCIL) has completed a peer review of the design in September 2003. BARC has
built several integral and separate effect test facilities for design validation of AHWR. In
view of the post Fukushima accident, several features have been incorporated, these include:
passive end shield and passive moderator cooling system; hard vent system; passive
autocatalytic re-combiners for hydrogen management; and core catcher. The new design
features are being validated in simulated test facilities. Site selection of AHWR has been
completed and the necessary clearance from competent authorities is underway. Preliminary
safety analysis report (PSAR) is ready for submission to regulatory body after obtaining site
clearances from competent authorities.
10. Plant Economics
The low capital cost and advanced safety features and of the AHWR could facilitate its easy
deployment. The main design features of the AHWR leading to reduced capital cost per
MW(e) are elimination of main coolant pump and associated equipment, substitution of SGs
with steam drums, elimination of heavy water recovery and tritium management systems andshop-assembled coolant channel assemblies. The reactor has two important provisions for
reducing O&M costs: (a) the elimination of potential for heavy water leakages from the main
coolant system such as in conventional PHWRs. This saves the recurring cost for heavy watermake up; (b) an easily replaceable coolant channel design.

52

IRIS (IRIS International Consortium)
1. Introduction
IRIS is an LWR with a modular, integral primary system configuration. The concept was
originally pursued by an international group of organizations led by Westinghouse. Currently
the IRIS related activities, especially those devoted to large scale integral testing, are being
pursued by Italian organisations. Its principle characteristics are:•medium power of up to 335 MW(e) per module;•
simplified compact design where the primary vessel houses the steam generators,pressurizer and pumps;•
an effective safety approach of active and passive safety systems; optimized
maintenance with intervals of at least four years.

Figure 1: Reactor System Configuration of IRIS
(Reproduced courtesy of POLIMI)
2. Target Application
The primary application of IRIS design is electricity production. However, this integrated
PWR can support the heat producing process and seawater desalination options. Couplingwith renewable energy parks and energy storage systems has been addressed as well.

53

3. Development Milestones
2001

Conceptual design completion

2001

Preliminary design start-up

2002

Pre-licensing process activities

4. General Design Description
Design Philosophy
IRIS is designed to satisfy enhanced safety, improved economics, proliferation resistance andwaste minimization.
Nuclear Steam Supply System
All the major NSSS equipment i.e. reactor coolant pumps, steam generators and pressurizer
are located inside the RPV, resulting in a more compact configuration and elimination of thelarge loss-of-coolant accident.
Reactor CoreThe IRIS core is an evolutionary design based on conventional UO2 fuel enriched to 4.95%.
An IRIS fuel assembly consists of 264 fuel rods with a 0.374 in. outer diameter in a 17×17
square array. The central position is reserved for in-core instrumentation, and 24 positions
have guide thimbles for the control rods. Low-power density is achieved by employing a coreconfiguration consisting of 89 fuel assemblies with a 14-ft (4.267m) active fuel height, and anominal thermal power of 1000 MW. The core is designed for a 3–3.5-year cycle with halfcore reload to optimize the overall fuel economics while maximizing the discharge burnup. In
addition, a 4-year straight burn fuel cycle can also be implemented to improve the overallplant availability, but at the expense of a somewhat reduced discharge burnup.
Reactivity Control
Reactivity control in IRIS is achieved through solid burnable absorbers, control rods, and theuse of a limited amount of soluble boron in the reactor coolant. The reduced use of solubleboron makes the moderator temperature coefficient more negative, thus increasing inherent
safety. Control rod drive mechanisms (CRDMs) are located inside the vessel, in the region
above the core and surrounded by the steam generators. Their advantages are in safety andoperation. Safety-wise, the uncontrolled rod ejection accident is eliminated because there isno potential 2000-psi differential pressure to drive out the CRDM extension shafts.
Operation-wise, the absence of CRDM nozzle penetrations in the upper head eliminates all
the operational problems related with corrosion cracking of these nozzle welds and seals.
Reactor Pressure Vessel and Internals
The IRIS reactor vessel houses not only the nuclear fuel and control rods, but also all the
major reactor coolant system components: eight small, spool type, reactor coolant pumps;
eight modular, helical coil, once through steam generators; a pressurizer located in the RV
upper head; the control rod drive mechanisms; and, a steel reflector which surrounds the coreand improves neutron economy, as well as it provides additional internal shielding. This
integral RV arrangement eliminates the individual component pressure vessels and large
connecting loop piping between them, resulting in a more compact configuration and in theelimination of the large loss-of-coolant accident as a design basis event. Because the IRIS
integral vessel contains all the RCS components, it is larger than the RV of a traditional looptype PWR. It has an internal diameter of 6.21 m and an overall height of 22.2 m including theclosure head.

54

Reactor Coolant System
The integral reactor coolant system of IRIS consists of 8 helical-coil steam generators and 8
spool type primary coolant pumps. The motor and pump consist of two concentric cylinders,
where the outer ring is the stationary stator and the inner ring is the rotor that carries highspecific speed pump impellers. The spool type pump is located entirely within the reactor
vessel, with only small penetrations for the electrical power cables and for water coolingsupply and return. Water flows upwards through the core and then through the riser region(defined by the extended core barrel). At the top of the riser, the coolant is directed into theupper part of the annular plenum between the extended core barrel and the RV inside wall,
where the suction of the reactor coolant pumps is located. The flow from each pump is
directed downward through its associated helical coil steam generator module. The primary
flow path continues down through the annular downcomer region outside the core to thelower plenum and then back to the core completing the circuit.
Steam Generator
The IRIS has once-through steam generators (OTSGs) with helical-coil tube bundle design
with the primary fluid outside the tubes. Eight steam generator modules are located in theannular space between the core barrel and the reactor vessel. Each IRIS SG module consistsof a central inner column which supports the tubes, the lower feed water header and the upper
steam header. The enveloping outer diameter of the tube bundle is 1.64 m. Each SG has 656
tubes, and the tubes and headers are designed for the full external RCS pressure. The tubes
are connected to the vertical sides of the lower feedwater header and the upper steam header.
The SG is supported from the RV wall and the headers are bolted to the vessel from the insideof the feed inlet and steam outlet pipes. The steam and feed lines and the emergency heatremoval system (EHRS) are designed for full primary pressure of 15.5 MPa. The EHRS does
not inject water, but only removes heat from the reactor via the SGs.
Pressurizer
The IRIS pressurizer is integrated into the upper head of the reactor vessel. The pressurizer
region is defined by an insulated, inverted top-hat structure that divides the circulating reactor
coolant flow path from the saturated pressurizer water. This structure includes a closed cell
insulation to minimize the heat transfer between the hotter pressurizer fluid and the subcooled
primary water. Annular heater rods are located in the bottom portion of the inverted top-hat
which contains holes to allow water insurge and outsurge to/from the pressurizer region.
These surge holes are located just below the heater rods so that insurge fluid flows up along
the heater elements. By utilizing the upper head region of the reactor vessel, the IRIS
pressurizer provides a very large water and steam volume, as compared to plants with a
traditional, separate, pressurizer vessel. The IRIS pressurizer has a total volume of ∼71 m3,
which includes a steam volume of 49 m3.

55

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

IRIS

Country of origin

International Consortium

Reactor type

Integral PWR

Electrical capacity (MW(e))

335

Thermal capacity (MW(th))

1000

Expected Capacity Factor (%)> 96

Design Life (years)

60

2Plant Footprint (m )14000 (four units layout)

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)15.5Core inlet/exit temperatures (oC)292/ 330

Main Reactivity Control Mechanism

ICRDM (Internally driven Control Rods)

RPV Height (m)21.3

RPV Diameter (m)6.2

Module Weight (metric ton)-

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine cycle

Fuel Type/Assembly Array

UO2/MOX/17x17 square

Fuel Assembly Active Length (m)4.26

Number of Fuel Assemblies

89

Fuel Enrichment (%)

4.95

Fuel Burnup (GWd/ton)

65 (max)

Fuel Cycle (months)

48 (max)

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

4

Refueling Outage (days)-

Distinguishing features

Integral primary system configuration

Modules per Plant

1 to 4

Target Construction Duration (months)

36

Seismic design (g)0.30

Predicted core damage frequency (per reactor year)

1E–8

Design Status

Basic design56

5. Safety Features
IRIS adopts passive safety systems and the safety by design philosophy including the risk
informed approach. Due to IRIS’s integral configuration, by design (i.e. with no interventionof either active or passive systems) a variety of accidents either are eliminated or theirconsequences and/or probability of occurring are greatly reduced. In fact, 88% of class IV
accidents (the ones with the possibility of radiation release) are either eliminated or
downgraded. This provides a defence in depth that allow IRIS eliminates emergency response
zone. The auxiliary building is fully seismically isolated. This provides a high level ofdefence in depth that may allow IRIS to claim no need for an emergency response zone. The
auxiliary building of IRIS is seismically isolated. The IRIS pressure suppression containment
vessel has a spherical configuration and it is 25 m in diameter. In case of small break loss of
coolant accident (SB LOCA), the RPV and containment become thermodynamically coupled.
The pressure differential across the break equalizes quickly and LOCA is stopped. The coreremains covered for all postulated breaks during the whole transient. The heat sink isdesigned to provide cooling for 7 days without operator action or off-site assistance forreplenishing.

Figure 2: SBLOCA safety strategy (Reproduced courtesy of POLIMI)
Engineered Safety System Approach and Configuration
IRIS has passive emergency heat removal system (EHRS) made of four independent
subsystems, each of which has a horizontal, U-tube heat exchanger connected to a separate
SG feed/steam line. These heat exchangers are immersed in the refueling water storage tank
(RWST) located outside the containment structure. The RWST water provides the heat sink tothe environment for the EHRS heat exchangers. The EHRS is sized so that a single subsystem
can provide core decay heat removal in the case of a loss of secondary system heat removal
capability. The EHRS operates in natural circulation, removing heat from the primary systemthrough the steam generators heat transfer surface, condensing the steam produced in the
EHRS heat exchanger, transferring the heat to the RWST water, and returning the condensateback to the SG. The EHRS provides both the main post-LOCA depressurization(depressurization without loss of mass) of the primary system and the core cooling functions.
It performs these functions by condensing the steam produced by the core directly inside thereactor vessel. This minimizes the break flow and actually reverses it for a portion of the
LOCA response, while transferring the decay heat to the environment. The safety strategy of
IRIS provides a diverse means of core shutdown by makeup of borated water from theemergency boration tanks (EBT) in addition to the control rods; also, the EHRS provides ameans of core cooling and heat removal to the environment in the event that normally57

available active systems are not available. In the event of a significant loss of primary-sidewater inventory, the primary line of defence for IRIS is represented by the large coolant
inventory in the reactor vessel and the fact that EHRS operation limits the loss of mass, thus
maintaining a sufficient inventory in the primary system and guaranteeing that the core willremain covered for all postulated events

EHRS
Heat Exchanger
SafetyValve

ADS/PORV(1 of 1)

SafetyValve

IntegralReactor
VesselMain Steam Line (1 of 4)
Isolation Valves

RCP(1 of 8)P/H

(1 0f 2)

Suppression
Pool (1 0f 6)

Steam Generator(1 of 8)FO

Suppression
Pool Gas
Space

SG
Make
up
Tank

FO

DVI

Long Term Core Makeup
from RV Cavity(1 of 2)

P/H

Emergency Heat Removal
System (EHRS)
1 of 4 Subsystems

SG Steam Lines(2 of 8)EBT

Refueling Water Storage
Tank (1 of 1)

AUX.
T.B.
BLDG.

SG FeedWater Lines(2 of 8)FO

FO

Main Feed Line (1 of 4)
Isolation Valves
P/H

P/H

RV Cavity

Start Up FeedWater

Figure 3: Engineered safety features of IRIS (Reproduced courtesy of POLIMI)
Containment system
Because the IRIS integral RV configuration eliminates the loop piping and the externallylocated steam generators, pumps and pressurizer with their individual vessels, the footprint of
the patent-pending IRIS containment system is greatly reduced. This size reduction,
combined with the spherical geometry, results in a design pressure capability at least threetimes higher than a typical loop reactor cylindrical containment, assuming the same metal
thickness and stress level in the shell. The current layout features a spherical, steel
containment vessel (CV) that is 25 m (82 ft.) in diameter as shown in Figure. 4. The CV isconstructed of 1-¾ in. steel plate and has a design pressure capability of 1.4 MPa (∼190psig). The CV has a bolted and flanged closure head at the top that provides access to the RV
upper head flange and bolting. Refueling of the reactor is accomplished by removing thecontainment vessel closure head, installing a sealing collar between the CV and RV, andremoving the RV head. The refueling cavity above the containment and RV is then flooded,
and the RV internals are removed and stored in the refueling cavity. Fuel assemblies arevertically lifted from the RV directly into a fuel handling and storage area, using a refuelingmachine located directly above the CV. Thus, no refueling equipment is required insidecontainment and the single refueling machine is used for all fuel movement activities. Figure
4 also shows the pressure suppression pool that limits the containment peak pressure to well58

below the CV design pressure. The suppression pool water is elevated such that it provides apotential source of elevated gravity driven makeup water to the RV. Also shown is the RV
flood-up cavity formed by the containment internal structure. The flood-up level is 9 m andensures that the lower section of the RV, where the core is located, is surrounded by water
following any postulated accident. The water flood-up height is sufficient to provide longterm gravity makeup, so that the RV water inventory is maintained above the core for anindefinite period of time. It also provides sufficient heat removal from the external RV
surface to prevent any vessel failure following beyond design basis scenarios.

Figure 4: IRIS spherical steel containment arrangement.
(Reproduced courtesy of POLIMI)
6. Plant safety and Operational Performances
The IRIS design provides for multiple levels of defense for accident mitigation, resulting in
extremely low core damage probabilities. In addition to the traditional DID levels (barriers,
redundancy, diversity, etc.) IRIS introduces a very basic level of DID, i.e., elimination by
design of accident initiators or reduction of their consequences/probability. Furthermore, adistinguishing characteristic of IRIS is its capability of operating with long cycles. Even
though the reference design features a two-batch, 3-year fuel cycle, selected on the basis ofease of licensing and US utilities preference, IRIS is capable of eventually operating instraight burn with a core lifetime of up to 8 years. However, the significant advantages
connected with a long refueling period in reducing O&M costs are lost if the reactor still has
to be shut down on a 18–24-month interval for routine maintenance and inspection. Thus,
first and foremost, the IRIS primary system components are designed to have very highreliability to decrease the incidence of equipment failures and reduce the frequency of
required inspections or repairs. Next, IRIS has been designed to extend the need forscheduled maintenance outages to at least 48 months. Because of the 4-year maintenance59

cycle capability, the capacity factor of IRIS is expected to comfortably satisfy and exceed the
95% target, and personnel requirements are expected to be significantly reduced. Bothconsiderations will result in decreased O&M costs.
7. Plant Arrangement
Almost half of the IRIS containment vessel is located below ground, thus leaving only about
15 m above the ground (i.e., several times less than the containment of a large LWR). This
very low profile makes IRIS an extremely difficult target for aircraft flying terrorists; inaddition, the IRIS containment is inconspicuously housed in and protected by the reactor
building. The cost of putting the entire reactor underground was evaluated; it was judged to
be prohibitive for a competitive entry to the power market and unnecessary since the IRIS
design characteristics are such to offer both an economic and very effective approach to thisproblem.
8. Design and Licensing Status
The IRIS team has completed the design of the large scale test facility, currently underconstruction, to prepare for future design certification. R&D activities in the field of design
economics, financial risk and SMR competitiveness are under way.

60

DMS (Hitachi-GE Nuclear Energy, Japan)
1. Introduction
DMS stands for Double MS: Modular Simplified and Medium Small Reactor. The designconcept is developed by Hitachi-GE Nuclear Energy under the sponsorship of the Japan
Atomic Power Company (JAPC). The DMS is a small-sized boiling water reactor (BWR) thatgenerates a capacity of 840 MW(th) or about 300 MW(e). The DMS reactor aims to optimize
the design according to the power output and achieve high economy by utilizing proventechnologies of existing reactors. The heat produced in the core is removed by natural
circulation of the coolant, thus eliminating the recirculation pumps and their driving powersources. This feature allows for a simplified and compact reactor pressure vessel (RPV) andcontainment. Due to the natural circulation feature, reactor internals and systems are also
simplified. The main features of the DMS reactor design are the miniaturization andsimplification of systems and equipment, integrated modulation of construction,
standardization of equipment layouts and effective use of proven technology. The factoryfabricated module reduces the construction period and also enables the modules to betransported to the site.

Figure 1: Reactor System Configuration of DMS(Reproduced courtesy of Hitachi-GE Nuclear Energy)
2. Target Application
A small-to-medium sized BWR is suitable for remote regions with less developed grids andinfrastructures. DMS design provides a nonelectric use of energy such as for district heating,mining (oil sand extraction/steam assisted gravity drainage) and desalination.
61

3. Development Milestones
2000-2004

Conceptual design

2014

Basic design (pre-licensing)

2017~

Design Review or Design Certification

2020~

Proposal to Customer or Commercial Bid

2030~

Commercial Operation

4. General Design Description
The main features of the DMS reactor include:• Simplified reactor core structures: Application of small sized natural circulation
reactor, the elimination of steam separator and the application of small sized RPV.
• Simplified systems and safety systems: Natural circulation eliminates the reactor
internal pumps and proper equipment configuration and simplification is achieved byintegrating systems/equipment. e.g., two main steam lines, the application of passive
system, common use of systems, simplified ECCS structure, etc. Active core injection
system including emergency generator is simplified by taking advantage of the large
coolant inventory in the RPV.
• Rationalized layout and construction method: By applying simplified systems andcompact primary containment vessel, the volume of building is reduced to a large
extent. The construction period is also reduced by applying large-scale modular
construction in which equipment, facilities and building structure are integrated.
Design Philosophy
The DMS is developed with the concept of high-economy small sized reactor of short
construction period to meet the diversified market needs. The design is based on a small sized
reactor pressure vessel, simplified safety systems, rationalized layout and architecturaldesign. The reactor is based on a proven technology with experience from existing BWRs,
intended for using systems and equipment that requires no large-scale development. The
DMS is designed to obtain a high safety level equivalent to the existing reactors with
optimized operations and maintenance performance in accordance with the level of output
power.
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) of the DMS is a direct cycle, where steam,
generated in the core, goes into a turbine directly. NSSS includes the reactor core andinternals, reactor pressure vessel and the coolant/steam piping within the containment system.
Primary pump, as well as primary re-circulation loop, is not equipped, due to the feature ofnatural circulation system. These system features contribute to remove a potential of LOCA.
Also, all of the large sized piping connected to the RPV is designed to be located above a topof an active fuel so that the DMS core is able to be covered by coolant even if the most severe
loss of coolant accident (LOCA) occurs. The NSSS of the DMS is designed to be able toremove core power by natural circulation core flow during normal operation, and the core canbe also cooled by natural circulation flow even if an anticipated transient or accident occurs.

62

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Hitachi-GE Nuclear EnergyCountry of origin

Japan

Reactor type

Boiling Water Reactor

Electrical capacity (MW(e))

About 300

Thermal capacity (MW(th))

840

Expected capacity factor> 87

Design life (years)

60

2Plant footprint (m )--Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure (MPa)7.17
oFeedwater /Core exit temperatures ( C)

186 / 287

Main reactivity control mechanism

Control rod drive, motor operated in normal operation.

RPV height (m)15

RPV diameter (m)4.8

Module weight (metric ton)

500~600 ton

Configuration of reactor coolant system

Integral type

Power conversion process

Direct Rankine cycle

Fuel type/assembly array

UO2 pellet/10x10 square configuration with channel box

Fuel assembly active length (m)2.0Number of fuel assemblies

400 (Short length fuel assembly)

Fuel enrichment (%)<5

Fuel burnup (GWd/ton)<60

Fuel cycle (months)

24

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (Passive + Active) system

Number of safety trains

3 active and 1 passive system

Refuelling outage (days)

30 (target)

Distinguishing features
Modules per plant

Simple reactor design, natural circulation system, hybrid
safety system, multipurpose use of energy.
1Target construction duration (months)~24

Seismic design

0.45gCore damage frequency (per reactor-year)

5.0E–8Design Status

Basic design

Reactor Core
The reactor core is loaded with 400 fuel bundles. The fuel active length is 2.0 m with theenrichment of less than 5 wt%. The short active fuel length reduces the pressure drop in coreand enables natural circulation. The short fuel length increases the number of fuel assemblies
necessary to secure the required thermal output increases, which results in increasing thediameter of RPV and the number of control rod drives, but the flow rate of natural circulation

63

can be reduced, making possible the reduction of RPV height. The core power density is
about 44 MW/m3. The core can produce energy with the refuelling period of 24 months.
Reactivity Control
The DMS has two kinds of diverse rector shutdown systems (i.e., control rod (CR)/control
rod driving system (CRD) and standby liquid control system (SLCS)). CR uses B4C or Hf asa neutron absorber and it is designed to be inserted from the bottom of the core. Every CR
has an independent CRD per each at the bottom of the reactor vessel. Since the CRD of DMS
has a motor-controlled fine motion capability, it is also called Fine motion CRD (FMCRD)
and it controls a positioning of CR, including insertion and withdrawal. DMS has 97 CRs and
FMCRDs. FMCRD has two kind of independent operation mode, one is fine motion control
by the electric motor, and the other is rapid motion (scram) by the hydraulic pressure. In thenormal operation, the reactivity in the core is controlled by the fine motion feature of
FMCRD.
Reactor Pressure Vessel and Internals
The size of the RPV is one of the main factors that determine the size of primary containment
vessel (PCV) and also greatly influence construction cost of the reactor building. The small
sized RPV in the DMS reactor is achieved by simplifying core internals through small sized
reactor core and natural circulation, eliminating steam separator. The flat core concept withshort-length and large core diameter is adopted to the DMS and the power density of the
DMS core is reduced to about 44MW/m3.The low power density results in a moderate
evaporation rate and lower steam velocity in the upper plenum of the RPV. This let the design
to adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator
and thus helps minimize the RPV and PCV sizes.
Reactor Coolant System
The DMS reactor primary cooling mechanism under normal operating condition andshutdown condition is by natural circulation of coolant. In order to enhance natural
circulation driving force, divided chimney of about 3m height is adopted above the core. The
reactor coolant system (RCS) is designed to ensure adequate cooling of reactor core under all
operational states, and during and following all postulated off normal conditions. A large
coolant inventory is achieved in the reactor due to the increased height and diameter of the
RPV. Like the conventional BWR, steam separation is performed inside the RPV. In DMS
however, this mechanism is done through free surface separation (FSS) in which the steam isseparated from water by gravity force. Hence, no physical separator assembly is required.
5. Safety Features
A large coolant inventory in achieved in the DMS compared with the forced circulation typereactor of the same output level, because the height of RPV is increased to secure the driving
power of natural circulation and the diameter of RPV is increased due to increased number offuel assemblies. These results in relieving the influence on reactor core in case of transientphenomena or medium and small piping break accident, so that high pressure core flooder(HPCF) system, which is equipped in a conventional BWR, can be eliminated. At the sametime, because the load of emergency generator is mostly shared by the motor capacity of
conventional HPCF, it becomes possible not only to eliminate HPCF system pumps but also
to reduce the capacity of emergency generator. By optimizing the configuration of systems inthis way, it becomes possible to simplify facilities to a large extent while keeping safety at thelevel equivalent to conventional reactors.

64

Engineered Safety System Approach and Configuration
As a defence-in-depth measure, enhanced hybrid safety systems that combine passive andactive methods are adopted. The safety system configuration of DMS has been rationallysimplified compared to a conventional large BWR (ABWR). There are four main distinctive
features: (1) High pressure core flooder (HPCF) equipped in conventional ABWR eliminated
due to the larger coolant inventory in the RPV of the DMS than ABWR; (2) Isolation
condenser (IC) and passive containment cooling system (PCCS) were added to the active
system as a countermeasure against long-term SBO. IC and PCCS can passively remove thedecay heat during at least 10 days; (3) Gas turbine generator (GTG) was adopted instead of
conventional diesel generator (D/G). GTG includes less auxiliary equipment than D/G, somaintenance load decreases and reliability increases. Though required time for start-up of
GTG is longer than that of D/G, DMS can adopt GTG because DMS has large time margin
until water level in the RPV reaches below the top of an active fuel; and (4) Reactor coreisolation cooling (RCIC) system and low pressure core flooder (LPFL) system wererationally integrated as Hybrid RCIC. RCIC can inject water into the RPV by using steam
generated in the RPV under high RPV pressure and LPFL can inject water by motor-driven
pump under low RPV pressure. The hybrid RCIC can inject water by using steam power
under high RPV pressure, and by using electric power under low RPV pressure. Long-term
SBO and design basis accident (DBA) were preliminary analyzed and it was confirmed thatthe core could be cooled for 10 days against SBO and peak cladding temperature (PCT) was
kept less than 1200oC even against the most severe DBA. It should be noted that optimizationof the safety system configuration continues under the discussion with potential customers
and safety regulators to meet their expectations
Apprication of passive systems

Moisture separator
High-pressure
turbinePrimary containment
vessel (PCV)

Low-pressure
turbineGenerator

Main Steam Lines
Feedwater lines
RPV

T urbine bypass line

Maincondenser
High-pressure
feedwater heater

Sea
water
Condensate pump

Low-pressure
feedwater heater

Control rod
drive
mechanism

T o maincondenser

Condensate filter

Main feedwater pump

Condensate demineralizer～

Apprication of GTG

Figure 2: Main configurations and safety features of DMS(Reproduced courtesy of Hitachi-GE Nuclear Energy)
Decay Heat Removal System
The residual heat removal system (RHR) includes a number of pumps and heat exchangers
which can be used to cool water in the RPV or the suppression pool (S/P) in the PCV. The
RHR can remove residual heat not only during normal shutdown and reactor servicing period
but also during an accident such as LOCA. The PCV cooling is accomplished by extracting65and cooling the S/P water, and injecting cooled water back to either the S/P injection line or
the containment spray lines. The IC and PCCS are equipped in the DMS as passive safetysystems without using any AC power. The IC and PCCS can condense steam generated in the
RPV and steam released from the RPV to the PCV via ruptured piping during LOCA,
respectively. The heat exchangers of the IC and PCCS can be cooled by the dedicated waterpool located above the PCV, which is filled with an amount of water enough to remove decay
heat at least 10 days passively.
Containment System
Steel containment is applied to achieve the design pressure (427 kPa[g]) equivalent to that ofMark-I type containment, and the quantity of material is reduced by reducing the diameterand height of PCV by adopting dish shape drywell and eccentric RPV arrangement. The
concept of compact PCV that improves constructability is established. The pressure
suppression containment that is experienced in BWR and ABWR was applied while
compactness was aimed at by eliminating steam separator, thus reducing the height of RPV
and the number of main steam pipes. The decrease in PCV height is achieved by reducing theactive fuel length of the DMS core, which is about 2 meters compared with 3.7 meters in theconventional BWR. The PCV is inserted by nitrogen during normal operation, therefore,hydrogen combustion in the PCV in early timing is practically eliminated. For a long-term
accident, a few number of passive autocatalytic re-combiners (PARs) are planned to equip in
the PCV to react hydrogen and oxygen generated due to water radiolysis.

Figure 3: Compact primary containment vessel of DMS(Reproduced courtesy of Hitachi-GE Nuclear Energy)
6. Plant safety and Operational Performances
The performance of the plant is improved by applying the main steam isolation valve of low
pressure loss developed for large sized reactors, for which the number of main steam pipes
are reduced to two in order to make the bore diameter of the pipes equivalent to that ofconventional BWR, as well as minimizing PCV by relieving the restrictions on the layout in66

primary containment vessel. In addition, considering the trade-offs of thermal efficiency andinstallation cost, facilities are optimized to realize the concept of simplifying facilities to alarge extent.
7. Instrumentation and Control systems
The digital I&C systems (e.g. microprocessors, field programmable gate arrays (FPGAs)) aremainly adopted in the DMS design and make use of fault detection and fault tolerance.
Diversity is also important particularly in providing a countermeasure against common cause
failure (CCF). Hardwired back-up safety system based on analogue technology is planned to
be installed to the DMS for diversity to mitigate influence of CCF of the digital I&C system.
The I&C system includes the safety system logic and control (SSLC), the plant control
systems, the hardwired back-up safety system, the auxiliary control system, and the plantcomputer system. The Reactor Protection system (RPS) which initiates rapid and automatic
shutdown (scram), Emergency Core Cooling system / Emergency Safety features(ECCS/ESF) which initiates ECCS are included in the SSLC.
8. Plant Arrangement
The concept of building layout and plant construction method is established, aiming at
reducing the quantity of construction material, shortening the construction period andstandardizing the plant design. The reduction in construction cost is established by applying
steel plate reinforced concrete structure, simplifying the shape of building and the building
structure.

Figure 4: Standardized reactor building(Reproduced courtesy of Hitachi-GE Nuclear Energy)
Reactor building
The reactor building is minimized by both system simplification and PCV compactness. The
number of system component is reduced by adoption of large capacity equipment, common
use of single equipment for different system, and adoption of passive system. PCV
compactness is achieved by dish shape drywell and eccentric RPV arrangement, i.e., the RPV67

is installed not at the real centre but at an eccentric centre of the PVC. Compact PCV lets thenumber of floor levels to reduce from six in current ABWR’s to four, which contribute tosaving in the construction period. The building is divided into fixed standard area, where
hardly influenced by site conditions and variable flexible area which may depend on site
conditions. The main power block surrounding the PCV or the secondary containment isdesigned to be the standard area. On the other hand, the circumferential area such as theelectrical room, plant make up facilities, etc. is designed to be flexible area. By this approachof rationalized layout, it is possible to realize the rate of building volume per unit output
power that is equivalent to ABWR.
Control building
The Control Building (C/B) includes the main control room (MCR), the computer facility, thecable tunnels, some of the plant switchgear, emergency batteries, reactor building water
system and a heating ventilating and air conditioning system (HVAC). During the plantnormal/emergency operation, the HVAC provides ventilation and air conditioning function.
For instance, in case of fire, smoke is discharged to outdoor and fresh air is supplied to the
C/B.
Balance of plant:
The advanced BOP system allows the utilization of produced heat for non-electricalapplications such as process heat, mining (oil sand extraction) and desalination.
Turbine Generator building
The structure of turbine systems are simplified to a large extent by applying the single
casing design that uses 41 inch turbine, which is proper for the output level as well as
by using single shell condenser and single train (4-stage heating) feed-water heater. The
number of equipment is reduced by integrating high and low pressure condensate
pumps and optimizing the configuration of systems.
Electric power systems
The DMS is planned to connect to the external grid via a main connection and astandby connection. The main connection is the connection between the generator
transformer (GT) and the external grid. The standby connection is the connection
between the auxiliary standby transformer (AST) and the external grid.
9. Design and Licensing Status
At the moment, there is no domestic license or pre-license activities for SMR in general,since there is no SMR construction project in Japan. Some SMR design applied or will apply
for a pre-licensing in USA or Canada (i.e., in customer’s country).
(Note: Japan does not have a process of “vendor’s design assessment”, such as DC in USA,GDA in UK or VDR in Canada. Vendor’s design is reviewed in the license process of theEstablishment Permit (Construction license) of the NPP construction project).
10. Plant Economics
The total cost per capacity ($/kW) of the DMS is estimated to be equivalent to that of aconventional ABWR.

68

IMR (Mitsubishi Heavy Industries, Japan)
1. Introduction
The Integrated Modular Water Reactor (IMR) is a medium sized power reactor with areference output of 1000 MW(th) producing an electrical output of 350 MW(e). The IMR isdeveloped for potential deployment after 2025. IMR employs the hybrid heat transportsystem (HHTS), a natural circulation system for primary heat transportation. The in-vessel
control rod drive mechanism (CRDM) is adopted as the primary means of reactivity control.
These design features allow the elimination of the emergency core cooling system (ECCS).

Figure 1: Reactor System Configuration of IMR(Reproduced courtesy of Mitsubishi Heavy Industries, Ltd.)
2. Target Application
The IMR is primarily designed as a land-based power station module to generate electricity.
The capacity of the power station can easily be uprated and adjusted to the demand by
installing additional modules. Because of its modular characteristics, it is suitable for largescale power stations consisting of several modules and also suitable for small distributedpower stations, especially when capacity of grids are small. IMR also has the capability for
district heating, seawater desalination, process steam production and so forth. IMR bases on
the equipment and systems that require no large-scale development and therefore can rapidly
be put to practical use in the market, including the period necessary to obtain authority’s
approval for the construction and operation of power plant.

69

3. Development Milestones
1999

IMR started its conceptual design study at Mitsubishi Heavy Industries (MHI).

2001 – 2004

An industry-university group led by MHI, including Kyoto University, Central Research Instituteof Electric Power Industries (CRIEPI), the Japan Atomic Power Company (JAPC), and MHI weredeveloping related key technologies through two projects, funded by the Japan Ministry of
Economy, Trade and Industry. In the first project, the feasibility of 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 study the startup ﬂow instability

2005 – 20072010 – 2013

4. General Design Description
Design Philosophy
The IMR is an integral type water reactor where the primary system components are all
installed inside the reactor pressure vessel (RPV). Its features are as follows:
•
•
•
•
•
•
•

Main coolant piping and primary coolant pumps are eliminated by adopting integrated
natural circulation system;
Pressurizer is eliminated by adopting the self-pressurization system;
The CRDMs are located inside the RPV;
There are two types of steam generators (SGs) inside the RPV. One is located in thevapour portion in the RPV (called SG in vapour (SGV)), and the other is located in
the liquid portion in the RPV (called SG in liquid (SGL));
SGs are also used as decay heat removal heat exchangers during accidents and normal
startup and shutdown operations;
A passive safety system, which does not require any external support (Stand-alone
direct heat removal system: SDHS) is adopted;
ECCSs are not required.

The IMR is a PWR with moderation ratios similar to those of operating PWRs, and thus, its
properties of fresh and spent fuel are also similar. This allows for the basic adoption of
conventional safeguards procedures and PWR management practices for new and spent fuel.
Simple support systems, such as the component cooling water system, the essential service
water system and the emergency AC power system, are designed as non-safety grade
systems, made possible by use of a stand-alone diesel generator.
Nuclear Steam Supply System
The HHTS is employed to transport the fission energy released in the fuel to the SGs by bothvapour formation and liquid temperature rise. The energy transported by vapour produces
secondary steam in SGV, and the energy transported by liquid temperature rise produces
secondary steam in SGL. The SGV also has a function of primary system pressure control,
and the SGL has the function of core power control through the core inlet temperature bycontrolling the feedwater flow rate.
Reactor CoreThe IMR has a core consisting of 97 fuel assemblies of 21×21 array with an averageenrichment of 4.95% and produces a thermal output of approximately 1000 MW(th). 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 the average discharged burnup
is about 46 GW d/t, which is approximately the same as current PWRs.

70

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Mitsubishi Heavy Industries, Ltd.

Country of origin

Japan

Reactor type

Integral PWR

Electrical capacity (MW(e))

350

Thermal capacity (MW(th))

1000

Expected capacity factor> 87

Design life (years)

60

2Plant footprint (m )-Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure (MPa)

15.51
o

Core inlet/exit temperatures ( C)

329 / 345

Main reactivity control mechanism

Control rod drive mechanism

RPV height (m)17

RPV diameter (m)6RPV or module weight (metric ton)-

Configuration of reactor coolant system

Integral type

Power conversion process

Indirect Rankine cycle

Fuel type/assembly array

UO2 pellet / 21 x 21 square

Fuel assembly active length (m)2.4

Number of fuel assemblies

97

Fuel enrichment (%)

4.8

Fuel burnup (GWd/ton)> 40

Fuel cycle (months)

26

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (Passive + Active) system

Number of safety trains

4

Refuelling outage (days)< 30

Distinguishing features

Integral PWR with natural circulation; employs two types invessel circular steam generators, one in liquid region, one invapour region.

Modules per plant

1

Target construction duration (months)< 42

Seismic design (g)0.3Core damage frequency (per reactor-year)

2.90E–7

Design Status

Conceptual design completed.

The cladding material employs Zr–Nb alloy to obtain integrity at a temperature of 345°C andover the long reactor lifetime. To maintain the core thermal margin and to achieve a long fuelcycle, the core power density is reduced to 1/3 - 1/2 of a conventional PWR. The designrefuelling interval is 3 years in 3 batches of fuel replacement. The fuel rod design is the sameas that for a conventional PWR.

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Reactivity Control
The chemical shim reactivity control is not adopted but both control rods which containenriched boron 10 and burnable absorbers control the whole reactivity. Control rods whose
neutron absorber is 90wt% enriched B4C perform the reactivity control, and a soluble acidboron 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, which is 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% locallyand less than 40% in the core to keep bubbly flow conditions. To reduce axial power peaking
caused by coolant boiling, the fuel consists of two parts: the upper part with higher
enrichment and the lower part with lower enrichment. Additionally, hollow annular pellets areused in the upper part fuel to reduce axial differences of burnup rate. Two types of in-vessel
CRDMs are adopted. One is motor driven CRDM, which is applied to the control bank. The
basic development of this CRDM has been conducted by the Japan Atomic Power ResearchInstitute (JAERI). This CRDM has the function of controlling reactivity during operation byfinely stepping the control rod position. The other is the hydraulic type CRDM. This CRDM
has the scram function and applied to the shut-down bank. The control rods connected to thisCRDM are moved by hydraulic force from the bottom position to the top, and then held byelectro-magnetic force. When the scram signal is initiated, the control rods are released andinserted 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 invessel SGs. The inside diameter of the lower part of the RPV is reduced to about 4 m in orderto reduce the cold-side water volume. In order to eliminate the necessity for the considerationof 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 partof 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 andpressurizer, it also reduces maintenance requirements. In addition, the use of the HHTS
concept makes it possible to reduce the size of the RPV. The HHTS is a kind of two-phasenatural circulation system. The coolant starts boiling in the upper part of the core, and twophase coolant keeps bubbly flow and flows up in the riser and is condensed and cooled by theSGs. This design approach increases coolant flow rate and thus reduces the required the RPVheight 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 Generators
There are two types of SGs adopted in the IMR. One is the SG in vapour (SGV), which islocated above the water level in the RPV. The energy transported by vapour formation
generates secondary steam through SGV. In addition, since the vapour in the RPV iscondensed by SGV, controlling the feedwater flow rate to SGV can control the RPV pressure.
The other is the steam generator in liquid (SGL), which is located in the water in the RPV.
The energy transported by liquid temperature rise generates secondary steam through SGL,and also, since the core inlet temperature can be controlled by the amount of heat removal
through SGL, the core power can be controlled by controlling the feedwater flow rate to72

SGL. By this method, the movement of the control rods for controlling reactor power will beminimized. In Figure 2 for SGL, a U-type tube bundle is adopted, since it is necessary tominimize pressure drops on both the primary and secondary sides to maintain good natural
circulation performance. A C-type steam generator is adopted for SGV, because of the goodspace utilization in the vapour part of the RPV.

Figure 2: Left: Steam Generator in Liquid Portion; Right: Steam Generator in Vapour
Portion (Reproduced courtesy of Mitsubishi Heavy Industries, Ltd.)
Pressurizer
The pressurizer is eliminated by adopting the self-pressurization system.
5. Safety Features
Engineered Safety System Approach and Configuration
Significant improvements have been achieved in the safety goals and safety philosophy of the
IMR. 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. Since the diameter of the pipes connected to the primary system(reactor vessel) is limited to less than 10 mm, the water level in the RPV can be maintained at
normal levels by water injection from the charging pumps. There are two trains of the SDHS.
Therefore, if a malfunction such as SG tube leakage were to occur, system functions would
be maintained. The capacity of chemical and volume control system (CVCS) is covered witheight 3/4-inch pipes which are connected to reactor vessel because of the small reactor
coolant inventory. No ECCS and containment cooling/spray system is installed. Safety
injection systems are eliminated by adopting the SDHS and by limiting the nozzle diameter
connected to the primary system. Containment spray system is eliminated by adopting the
SDHS. The auxiliary feedwater system is used for startup and shutdown procedures during
normal operation. It is also used at the initial stage of the transients caused by the secondary
systems. When the plant safety is threatened only by the auxiliary feedwater system, thesafety system (SDHS) is actuated. The auxiliary feedwater system is not a safety system,
because the plant safety is guaranteed by SDHS. The IMR adopts simplified support systems,
such as the component cooling water system (CCWS), the essential service water system(ESWS) and the emergency AC power system. These are designed as non-safety grade73systems, possible by use of a stand-alone diesel generator. A simplified CVCS and waste
disposal system (WDS) is achieved by the boric-acid free design.
Decay Heat Removal System
For the mitigation system, the SDHS is activated. Figure 3 shows the concept of SDHS. In
this system, the decay heat is removed directly from inside the RPV to the atmosphere.
Therefore, even if water leakage occurred and the charging pumps fails to operate, waterleakage would be terminated automatically when the pressures inside and outside the RPV
are equalized. In the passive steam generator cooler (PSGC), decay heat is removed by watercooling in the early stage of the accident and then, the heat transfer mode is graduallyreplaced by air-cooling. Therefore, external support such as water, power and operators are
not necessary for maintaining plant safety.

Figure 3: Stand-alone Direct Heat Removal System of IMR(Reproduced courtesy of Mitsubishi Heavy Industries, Ltd.)
Containment System
A small containment vessel (CV) is made possible in the IMR by adopting an integrated
primary system design and simplified systems. The IMR adopts reinforced concrete
containment. The design pressure of the containment is planned to be 9 MPa to meet thesafety requirement that water leakage from RPV will be terminated automatically. Since thisCV is about one size larger than the RPV, it will be able to resist high pressure. The
containment is part of the reactor containment facility. The reactor containment facility is partof the engineered safety systems, which include SDHS. The containment system is designed
to suppress or prevent the possible dispersion of large quantities of radioactive materials,
which would be released if extensive fuel failures were to occur in the reactor resulting fromdamage or failure of the reactor facilities such as the primary cooling system, main steam
system, and feed water system.

74

Figure 4: Compact primary containment vessel of IMR(Reproduced courtesy of Mitsubishi Heavy Industries, Ltd.)6. Plant safety and Operational Performances
The IMR reactor is designed to be operated automatically within the range of 20 to 100% of
rated output power by the reactor control system. Even in the low output range below 20%,the control system can control the reactor automatically in the low power-operating mode.
Usually, the primary system pressure and reactor power are controlled feedwater and control
rods. When the load to the plant is changed, the feedwater flow rate and control rods arecontrolled simultaneously to steadily maintain the plant. The reactor control system isdesigned so that it can allow the following load demand change without causing a reactor
trip:
•
•
•

A +/- 10% step load demand change (within the range of 20 to 100%);
A +/- 5% per min ramp load demand change (within the range of 20 to 100%);
100% load rejection.

7. Plant Arrangement
The concept of building layout and plant construction method is established, aiming at
reducing the quantity of construction material, shortening the construction period andstandardizing the plant design. The reduction in construction cost is established by applying
steel plate reinforced concrete structure, simplifying the shape of building and the building
structure.

75

Figure 5: IMR Plant Layout Arrangement(Reproduced courtesy of Mitsubishi Heavy Industries, Ltd.)
Reactor building
Ground level is assumed at above sea level, and flat land is assumed. The bedrock is assumed
to be less than 40 meters below ground to enable the use of pile foundations. The reactor
building is adopted in an isolated building and the turbine building is not adopted. The reactor
building configuration is as follows:• Integrated building consisting of two units;• Exclusion of waste disposal facilities in another building;• Adoption of steel structures in the building considering the module method.
Balance of plant:
The advanced BOP system allows the utilization of produced heat for non-electricalapplications such as process heat, mining (oil sand extraction) and desalination.
Turbine Generator building
The turbine generator, turbine, condenser, moisture separator and reheater (MSR) andtheir auxiliary equipment are installed in the turbine building. The turbine generator isarranged with its axis in line with the reactor. Suitable spaces have been provided forinspection access, the transportation of tools for inspections and maintenance, anddisassembly in such a way that the volume of the building is reduced.
8. Design and Licensing Status
At the moment, there is no domestic license or pre-license activities for SMR in general,since there is no SMR construction project in Japan. The IMR conceptual design study was
initiated in 1999 by Mitsubishi Heavy Industries (MHI). A group led by MHI and 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 for components and design methods, and basic
design development are required before licensing.

76

SMART (KAERI, Republic of Korea)
1. Introduction
The System-integrated Modular Advanced ReacTor (SMART) is an integral PWR with a
rated electrical power of 100 MW(e) from 330 MW(th). SMART adopts advanced design
features to enhance the safety, reliability and economics. The advanced design features andtechnologies implemented in the SMART 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 the economics through system simplification, component modularization,
reduction of construction time and high plant availability. The advanced design features of
SMART include integrated steam generators (SG), in-vessel pressurizer and the horizontally
mounted canned motor pumps. Figure 1 shows SMART reactor system configuration.

Figure 1: Reactor system configuration of SMART (Reproduced courtesy of KAERI)
2. Target Applications
SMART is a multi-purpose application reactor for electricity production, sea water
desalination, district heating, process heat for industries and suitable for small or isolated
grids. SMART has a unit output enough to meet the demands of electricity and fresh water fora city of 100000 populations. The SMART multi-purpose application configuration is shownin Figure 2.

77

SMART

Electricity
SteamTransformer
Steam
Intake Facilities
M a in Eje ctor
Ste a m Supply
Ve nt Eje c tor

Sc a le Inhiv itor

SeawaterTo Outfa ll

Fe e d Wa ter

Fe e dSe a wa te r

1 s t Effe c t

Conde ns a te
Pum p

2 nd Effe c t

3 rd Effe c t

La s t Effe c t

Fina lConde ns e r

Fresh water

Fla s hConde ns e r

Desalination Plant

Brine Blowdown
Pum p

Dis tilla te
Pum p

Figure 2: SMART multi-purpose application configuration (Reproduced courtesy of KAERI)
3. Development Milestones
March 1999

Conceptual design development

March 2002

Basic design development

June 2007

SMART-PPS (Pre-Project Service)

July 2012

Technology verification, Standard Design Approval (SDA)

March 2012

First step of Post-Fukushima corrections and commercialization
Pre-project engineering agreement signed between Republic of Korea and Kingdomof Saudi Arabia for the deployment of SMART in the Gulf country

September 2015

4. General Design Description
Design Philosophy
SMART design adopts an integrated primary system, modularization and advanced passive
safety system to improve the safety, reliability as well as the economics. Safety performanceof SMART is assured by adopting passive safety systems together with severe accident
mitigation features. The passive safety features rely on gravity and natural circulation andrequire no active controls and operator intervention to cope with malfunctions and safety
events. Improvement in the economics is achieved through a system simplification, in-factoryfabrication, reduction of construction time and high plant availability.
Nuclear Steam Supply System
SMART has an integral reactor coolant system configuration that enables the elimination oflarge bore pipe connections resulting in the removal of the large break loss of coolant
accident (LOCA) from the design bases events. The nuclear steam supply system (NSSS)
consists of the reactor core, helically coiled steam generators and a pressurizer integrated in
the reactor pressure vessel (RPV). The primary cooling system is based on forced circulation
during normal operation. The system has natural circulation capability for emergency
condition.

78

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

KAERI

Country of origin

Republic of Korea

Reactor type

Integral PWR

Electrical capacity (MW(e))

100Thermal capacity (MW(th))

330

Expected Availability Factor (%)> 90

Design Life (years)

60

Plant Footprint (m2)90000

Coolant/moderator

Light water

Primary circulation

Forced circulation

Number of reactor coolant pump

4 (horizontally mounted on the reactor vessel)

System pressure (MPa)15oCore inlet/exit temperatures ( C)296/323

Main Reactivity Control Mechanism

Control rod driving mechanisms and soluble boron

RPV Height (m)18.5

RPV Diameter (m)6.5

Module Weight (metric ton)

1070 (including coolant)

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine Cycle

Fuel Type/Assembly Array

UO2 pellet/17x17 square

Fuel Assembly Active Length (m)2Number of Fuel Assemblies

57

Fuel Enrichment (%)<5

Fuel Burnup (GWd/ton)< 60

Fuel Cycle (months)

36

Cogeneration Capability

Yes

Approach to Engineered Safety Systems

Passive

Number of safety trains

4

Refuelling Outage (days)

45

Distinguishing features

Coupling with desalination and process heat application, integrated
primary system, passive safety system.

Modules per Plant

1

Target Construction Duration (months)

36

Seismic design> 0.18 g automatic shutdown

Predicted core damage frequency per reactor year

2E-7(internal events)

Design Status

Licensed/Certified (Standard Design Approval)

Reactor Core
The low power density design with a slightly enriched UO2 fuelled core ensures a thermal
margin of greater than 15%, which can accommodate any anticipated transient event. Thisfeature ensures the core thermal reliability under normal and accident conditions. The fuelassembly (FA) is 2 m long and formed by a standard 17x17 square of UO2 ceramic fuel withthe enrichment of less than 5% similar to standard PWR fuel. There are 57 FAs in the core.
79

SMART fuel management is designed to achieve a maximum cycle length between refuellingoutages. A two-batch refuelling scheme without reprocessing provides a cycle of 1000
effective full power days for 36 months of operation. This reload scheme minimizes
complicated fuel shuffle schemes and enhances fuel utilization. The integral arrangement ofreactor components prevents easy access to the fuel as it would require a special equipment toload and unload the fuel in the reactor core.
Reactivity Control
Reactivity control during normal operation is achieved using control rods and soluble boron.
Burnable poison rods are introduced for flat 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 fuel type are selected so that reactivity of each assembly can be as flatas possible. SMART adopts typical magnetic-jack type control rod drive mechanism which
has been widely used in the commercial nuclear power plants (NPPs). The small number offuel assembly in the SMART assures a relatively high control rod worth.
Reactor Pressure Vessel and Internals
SMART integrated design means that the RPV contains all of the primary components. The
RPV houses the core structures, eight (8) helical once-through steam generators (SGs),
internal steam pressurizer and four (4) canned motor reactor coolant pumps.
Reactor Coolant System
SMART primary cooling system is based on forced circulation. The cooling of the reactor
core is achieved by the four (4) canned reactor coolant pumps (RCPs) horizontally mounted
at the upper shell of the RPV that circulate the coolant from the core to the SGs. Each RCP isan integral unit consisting of a canned asynchronous three phase motor and a mixed-flow,
single-stage pump. The RCPs actively recirculate the coolant routing it into the core after
having transferred heat to the secondary side of the SGs. The canned motor RCPs inherently
eliminate potential small break LOCA associated with a pump seal failure which is one of thedesign basis events for conventional reactors. The reactor coolant forced by RCPs flows
upward through the core and enters the shell side of the SG at the top of the SG. The
secondary side feedwater enters the helically coiled tube side at the bottom of the SG andflows upward to remove the heat from the shell side eventually exiting the SG in asuperheated steam condition. A large volume of the primary coolant provides large thermal
inertia and a long response time, and thus enhances the resistance to system transients andaccidents. These features ensure core thermal reliability under normal operation and any
design basis events.
Steam Generator
The SMART has eight (8) modular type once-through SGs with helically coiled tubes to
produce superheated steam under normal operating conditions. The SGs are located at thecircumferential periphery between the core support barrel and RPV above the core to provide
a driving force for natural circulation flow. The small inventory of the secondary side (tubeside) water in each SG prohibits a return to power following a steam line break accident. In
the case of accidents, the SG can be used as the heat exchanger for passive residual heatremoval system (PRHRS), which permits an independent operation of the PRHRS regardlessof the hydraulic condition of the primary system.
Pressurizer
The in-vessel pressurizer is designed to control the system pressure at nearly constant level
for the overall plant operation. The large free volume formed in the upper part of the RPV is

80

used as the pressurizer region. The primary system pressure is maintained nearly constant 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 for the load manoeuvring
operation. The reactor over-pressure at the postulated design basis accidents related with acontrol failure can be reduced through the actuation of the pressurizer safety valve (PSV).
5. Safety Features
Safety systems of SMART are designed to function automatically on demand. These consistof a reactor shutdown system, a passive safety injection system (PSIS), PRHRS, a shutdown
cooling system and a passive containment cooling system (PCCS). Additional safety systems
include reactor overpressure protection system such as automatic depressurization system(ADS) and pressurizer safety valves, and severe accident mitigation system. As a result thereactor safety is enhanced substantially and the core damage frequency is also reduced.
Engineered Safety System Approach and Configuration
SMART has passive safety features to cope with abnormal operating occurrences andpostulated design basis accidents, and severe accidents with core melts. The passive safetysystem is being developed to maintain a safe shutdown condition following design basis
accidents such as LOCA and non-LOCA transient events without AC power or operator
actions. The passive safety system consists mainly of PSIS, PRHRS, PCCS, and automatic
depressurization system (ADS). All the active safety features are being substituted with
passive versions, eliminating the necessity for emergency diesel generator (EDG) or operator
actions for at least 72 hours period. Engineered safety system of the SMART is shown in
Figure 3.

Figure 3: Engineered safety system of SMART (Reproduced courtesy of KAERI)
Decay Heat Removal System
After the reactor is shutdown, when the normal decay heat removal mechanism utilizing thesecondary system is not operable by any reason, the PRHRS brings the RCS to the safe
shutdown condition within 36 hours after accident initiation and maintains the safe shutdown
condition for at least another 36 hours. Therefore the safety function operates for 72 hours81without any corrective action by operator for the postulated design basis accidents. The safetyfunction of PRHRS is maintained continuously for a long term period when the emergency
cooling tank (ECT) is replenished periodically. The PRHRS consists of four independent
trains with a 33% capacity each, and each train is composed of an ECT, a heat exchanger anda makeup tank. In the design of PRHRS the possibility of loss of one train by a single failure
is eliminated. Each train of PRHRS has a pair of check valves and isolation valves, both ofwhich are installed on parallel lines. Therefore even if one valve fails, the whole train is stillin operation. To remove the possibility of common mode failure, diversity of actuator isprovided. Two kinds of different isolation valves are adopted, which are air-operated andelectro-hydraulic valves. Therefore, single failure is no longer an issue in the design of
PRHRS and three out-of-four trains are enough to remove the residual heat after an accident
occurs.
Emergency Core Cooling System
Passive safety injection system (SIS) provides emergency core cooling following postulated
design basis accidents. Emergency core cooling is performed using the four (4) core make-up
tanks (CMTs) and four (4) safety injection tanks (SITs). Core cooling inventory is maintained
through passive safety injection of CMTs and SITs with a 33% capacity each. The four (4)CMTs which are full of borated water provide makeup and borating functions to the RCS
during early stage of SBLOCA or non-LOCA event. The top and bottom of CMT areconnected to the RCS through the pressure balance line and the safety injection line,
respectively. Each safety injection line is isolated by 2 x 2 parallel closed valves, whichmeets single failure criteria. Each pressure balance line is normally open to maintain pressureof the CMT at RCS condition. This arrangement enables the CMT to inject water to the RCS
by gravity when the isolation valves are open. The isolation valves of the CMT injection line
are actuated to open by low pressurizer pressure signal, high containment pressure signal, or
PRHRS actuation signal. The four (4) SITs are filled with borated water and nitrogen gas at
atmospheric pressure. These tanks provide makeup water to the RCS at low pressure
conditions for about 72 hours in case of LOCA. The arrangement of the SIT is similar to thatof CMT except that the isolation valves are installed at the pressure balancing line. The
isolation valves of the SIT injection line are open when the RCS pressure reaches the set
point pressure.
Containment System
The containment system is designed to contain the radioactive fission products within thecontainment building and to protect the environment against primary coolant leakage. The
cylindrical primary containment system has a diameter of 44 m and a design pressure of 0.34
MPa. It protects the entire reactor systems from any external air-borne collisions. The
containment system includes twelve (12) passive autocatalytic hydrogen recombiners andfour (4) trains of passive containment cooling system (PCCS). The PCCS reduces
containment pressure and temperature in case of main steam line break (MSLB) or LOCA,
and removes fission products from the containment atmosphere following LOCA. The PCCS
consists of four (4) independent trains with a 33% capacity each, and each train is composedof a heat exchanger, connecting pipe and valves. The PCCS dissipates the heat from thecontainment atmosphere to the environment through the heat exchanger by natural
circulation.

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6. Plant safety and Operational Performances
The load follow operation of SMART is simpler than that of large PWR because only thesingle bank movement and small insertion limit is required. SMART is suitable for load
follow operation because of the small reactivity change for the power change due to theminimized coolant temperature change, relatively high lead bank worth due to a small
number of fuel assemblies and the short effective core height leading to rapid damping of
xenon oscillation. Daily load follow capability of SMART core performed shows that radial
peaking factor, 3-dimensional peaking factor and the axial offset were satisfied within design
limit. The load follow operation can be performed by the lead bank movement only withinbank insertion limit without the change of soluble boron.
7. Instrumentation and Control Systems
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, andsensing signal sharing for protection & control system. 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 signals from safety channels of the ex-coreneutron flux monitoring system are shared to use for reactivity control system. In–core
instrumentation system consists of 29 detector assemblies which are developed as mini typefor SMART with four stacked rhodium self–powered detectors, one background detector andone set of K type thermocouple for each.
8. Plant Arrangement
Plant main building consists of the reactor containment building, the auxiliary building, thecompound building (CPB), the emergency diesel generator building and the turbine-generator
building (TGB). The CPB consists of an access control area, a radwaste treatment area, and a
hot machine shop. For efficient radiation management, the plant main building is sub-divided
into two zones; the duty zone and the clean zone.
Reactor Building
The reactor building and fuel storage area are equipped with a full monitoring system withclosed circuit monitoring system (CCTV) to oversee and prevent unauthorized access to thefuel. Reactor building is a pre-stressed concrete shell structure composed of a right cylinder
with a hemispherical dome, with steel plate lining to act as a leak tight membrane. Reactor
building is erected on a common base-mat together with the auxiliary building in which themain control room and fuel storage area are located.
Main Control RoomThe SMART compact control room is designed for one man operation under normal
conditions of the plant. 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 thetasks according to the proper operating procedures. To achieve these goals, human factors
engineering (HFE) process and principles are applied and verified using the full scope
dynamic mock up. The four types of compact workstations such as safe shutdown monitoring
and control workstation, which also contains a set of hardwired switches for manual actuationof ESF equipment and reactor trip, main monitoring and control workstation, auxiliary
monitoring and control workstation, minimum inventory control workstation and a large
display panel are installed in the MCR.

83

Figure 4: SMART general plant arrangement with a single and a twin unit
(Reproduced courtesy of KAERI)
Balance of Plant:
Turbine Generator Building
The reference concept of the turbine plant has been developed including a couplingsystem for seawater desalination. The secondary system receives superheated steam fromthe NSSS. It uses most of the steam for electricity generation and preheaters, and theremainder for non-electric applications. Sea water desalination system may be used inconjunction with the secondary system. The steam transformer produces the motive
steam using steam extracted from a turbine and supplies it to the desalination plant.
Electric Power Systems
These systems include the main generator, main transformer, unit auxiliary transformers,
stand-by auxiliary transformers, diesel generators, and batteries. The electrical systems,
including the class 1E and non-class 1E, are based on a two train approach. The main
generator is connected to the grid via the main transformer with auxiliary transformers
connected between them. The unit auxiliary and/or the stand-by transformers supply
electrical power for start-up, normal operation and shutdown.
9. Design and Licensing status
Korea Atomic Energy Research Institute (KAERI) received the standard design approval
from the country’s nuclear safety commission in July 2012. In September 2015, a pre-projectengineering agreement was signed between the Republic of Korea and the Kingdom of Saudi
Arabia for deployment of SMART. A construction period of less than three (3) years fromfirst concrete to fuel load is predicted. A safety enhancement program to adopt passive safetysystem in SMART began in March 2012, and the testing and verification of the PRHRS and
PSIS were completed in the end of 2015.
10. Plant Economics
The target plant construction cost is $10000/kWe and an operating & maintenance cost of
4.5¢/kWh. For a twin unit of SMART, the total cost is expected to be 20% lesser.

84

UNITHERM (NIKIET, Russian Federation)
1. Introduction
The UNITHERM is a small transportable nuclear power plant (NPP) with a capacity of 30
MW(th) and a rated electrical output of 6.6 MW. UNITHERM is developed based upon
NIKIET’s experience in designing marine nuclear installations. The UNITHERM reactor isintended 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 andoperational experience of the WWER type reactors. The design aims for fabrication,assembly and commissioning of the NPP modules to be carried out at factory. The
UNITHERM reactor is designed to operate for 20-25 years without refuelling. The
UNITHERM reactor design allows for its use both as a land-based and barge mounted NPP.
NPP with UNITHERM may consist of a number of units depending on the purpose anddemand of costumers need.

Figure 1: Reactor System Configuration of UNITHERM(Reproduced courtesy of NIKIET)85

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, theconfiguration and design of the UNITHERM is sufficiently flexible to be adjusted or
modified for different target functions and user requirements, without compromising theunderlying principles of the design.
3. Development Milestones
1994

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

2012

Technical proposal on the UNITHERM reactor facility (WDR stage)

2015

Technical proposal for a SMR plant based on the UNITHERM reactor

4. General Design Description
The UNITHERM plant is designed to operate without presence of operating personnel. The
design aims to provide the reactor maintenance in routine and urgent cases from regional
centres common to several plants of this kind. The design adopts inherent safety features
which include: large coolant inventory, negative reactivity coefficients and a high flow rate ofcoolant in natural circulation. One of the main features of the UNITHERM is the suitability
for shop fabrication and testing.
Design Philosophy
NPPs with the UNITHERM reactor are designed for siting in remote regions with lessdeveloped 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 2025 years. The refuelling of the core will not be required during the plant service life.
Nuclear Steam Supply System
Primary circuit system is intended for heat removal from the reactor core and heat transfer tothe intermediate circuit fluid inside the intermediate heat exchanger. The system consists of amain circulation train and a pressurizing system. The natural circulation of primary coolant
takes place in the primary circuit. The intermediate circuit system is intended for heat transferfrom the intermediate circuit coolant to the secondary coolant (consumer’s circuit) inside asteam generator (SG). This system provides an additional localizing safety barrier to protect
the heat consumers against the ionizing radiation from radionuclides generated by primarycoolant 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 turbogenerator pumps) by means of heat transfer from the intermediate circuit
coolant inside the steam generator. Secondary coolant circulates by means of naturalconvection.

86

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer
Country of origin

NIKIET, Research and Development Institute of Poweri
i
RussianFederation

Reactor type

PWR

Electrical capacity (MW(e))

6.6

Thermal capacity (MW(th))

30

Expected capacity factor (%)

70

Design life (years)

25

2

Plant footprint (m )~ 10000

Coolant/moderator

High purity water

Primary circulation

Natural circulation

System pressure (MPa)16.5oCore inlet/exit temperatures ( C)

249 / 330

Main reactivity control mechanism

Soluble boron and control rod insertion

RPV height (m)9.8

RPV diameter (m)2.9

RPV or module weight (metric ton)180 (integral reactor module)

Configuration of reactor coolant system

Integral type

Power conversion process

Direct Rankine Cycle

Fuel type/assembly array
Fuel assembly active length (m)UO2 particles in a metallic (silumin or zirconium) matrix , metalceramic/ 54-55
1.1

Number of fuel assemblies

265

Fuel enrichment (%)

19.75

Fuel burnup (GWd/ton)

1.15

Fuel cycle (months)

200

Cogeneration capability

Yes (heat mode power: 2.5 MW(e), 20 MW(th) heat)Approach to engineered safety systems

Hybrid (Passive + Active) system

Number of safety trains

2

Refuelling outage (days)

45-50

Distinguishing features

Permanently operational autonomous passive system for abstract ofheat from reactor; guard vessel; iron-water biological protection ofreactor; passive systems for abstract of heat from the guard vessel
and biological shielding tanks

Modules per plant

As per customer requirement

Target construction duration (months)

84

Seismic design

VIII-IX-MSK 64

Core damage frequency (per reactor year)

1E–6

Design Status

Conceptual design87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 isdesigned as a cylindrical rod with four spacing ribs on its outer surface. The fuel is in theform 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 filled with silumin. A fuelelement 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 is characterized by ahigh 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 isfilled with the matrix composition.
Reactivity Control
The control element drive mechanisms (CEDMs) are designed to provide secure insertion ofrods in the core by gravity for reactivity control. Locking devices are installed in the CEDM
to avoid unauthorized withdrawal of control rods. Burnable absorbers are used to compensate
the decrease of reactivity due to fuel burn-up, temperature effect and by motion of thereactivity control rods during periodic maintenance.

Figure 2: UNITHERM schematic diagram(Reproduced courtesy of NIKIET)
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).

88

Reactor Coolant System
The UNITHERM primary cooling mechanism under normal operating condition andshutdown condition is by natural circulation of coolant. The heat energy released from thereactor 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 coreto the intermediate circuit coolant occurs in the built-in once-through intermediate heatexchanger and heat transfer from the intermediate coolant to the consumer’s circuit coolant
inside the SG. Both heat exchangers are made from titanium alloy. The intermediate heatexchanger 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 the environment under normal and design basis accidents is well below thelimits 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 potential hazardous activities related to the core refuelling, as thereactor core refuelling will not be required in the plant service life. This further simplifies theoperating technologies and enhances the proliferation resistance.
Engineered Safety System Approach and Configuration
The UNITHERM safety systems are based upon redundancy, diversity and the maximum useof the fail-safe systems. The UNITHERM employs passive safety systems and devices which
do not require actuation (such as containment, independent heat removal system, etc.) or canbe passively actuated (such as primary circuit systems and containment depressurization
system). The reliability and safety of the UNITHERM reactor is significantly improved dueto 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 containment is 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 andmeasures 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 tothe containment. The emergency core protection system is activated in response to the signals
from pressure transducers. Coolant leakage continues until the pressure values in the reactor
and containment are equalized. The remaining coolant inventory in the reactor is sufficient tomaintain circulation in the primary coolant circuit. The reactor is passively cooled via the89intermediate circuit and the independent heat removal circuit, whereas the containment heatis removed by the component cooling system. Additionally, an active user circuit withfeedwater supplied to the SG and steam-water mixture maybe utilized to increase the coolingrate. The iron-water biological shield acts as a system of 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.

Figure 3: (1) iron-water shielding tank; (2) containment; (3) shock-proof casing;(4) cooldown system heat exchanger; (5) safeguard vessel; and (6) the reactor(Reproduced courtesy of NIKIET)
Containment system
The integral reactor for land-based deployment is placed inside the leak-tight containment,
which is located within the concrete shock-resistant structure together with the biologicalshielding and reactor unit components. This structure enhances physical protection of thereactor unit from external impacts such as airplane crash, hurricane, tsunami, unauthorizedaccess, etc. The containment system is capable of maintaining the primary coolant circulation
as well as provides reactor cooldown and retention of radioactive products under the loss ofprimary circuit leak tightness. Passive safety systems for the removal of heat from thecontainment and biological shielding tanks are employed.
6. Plant Arrangement
Turbine Generator building
Turbine-generator assembly for UNITHERM NPP depends on the plant capacity andoperation mode requested by its users. The turbine operates using dry saturated steam in themode of steam outlet backpressure. With consideration of the continuous transfer of 5% heatto 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
90exhaust. An electric generator with an output of 6.6 MW(e) in combination with a singlephase intermediate circuit allows to obtain a superheated steam temperature of 285°С under
1.35 MPa.

Figure 4: The UNITHERM plant arrangement (Reproduced courtesy of NIKIET)

1 – reactor plant; 2 – reactor building; 3 – bridge crane; 4 – transfer lock; 5 – overpass with thecrane; 6 – pipelines and components inside the reactor building; 7 – turbine generator;
8 – turbine building; 9 – air cooler; 10 – administrative and amenity building; 11 – roads;
12 – site area

7. Design and Licensing Status
Based on the experience of NIKIET and other Russian institutions and enterprises in thedevelopment of marine nuclear installations, the UNITHERM NPP may require no major
research and technology development activities for its deployment. Once an agreement withthe user is reached and the technical assignment 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 and othercomponents.
8. Plant Economics
The serial production of plant equipment fully under shop conditions, including equipment
assembly and testing is expected to facilitate the low capital costs for plant construction

91

KARAT-45 (NIKIET, Russian Federation)
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 the base facility for the economic and social development of the
Arctic region and remote extreme northern areas of Russian Federation.

Figure 1: Reactor System Configuration of KARAT-45(Reproduced courtesy of NIKIET)
2. Target Application
KARAT-45 is a small boiling water reactor (BWR) with a rated power of 45 MW(e) designed
as an independent cogeneration plant for supply of electric power, steam heat and hot water,
as shown in Figure 2. KARAT-45 power unit has a high load follow capability to cope withdaily power variation from 20% to 100% of nominal capacity.

92

Figure 2: Application Scheme of KARAT-45(Reproduced courtesy of NIKIET)
3. Development Milestones
2015

Development of a Detailed Project Report

2016

Project definition, technical requirements, R&D program

2017 – 2019

Basic design, R&D, development of PSAR, expert review and licensing

2020 – 2022

Detailed design, fabrication of equipment, construction

2023

Operation license, first criticality, first start, commissioning

4. General Design Description
Design Philosophy
The BWR technology was selected as a basis for the design and technology development of
KARAT-45 due to the following rationales: BWR employs single circuit removal of heat so
capital cost for construction can be minimized; lower system pressure poses fewer challenges
to the reactor vessel; BWR has inherent self-protection and self-control properties due tonegative void and temperature reactivity coefficients. KARAT-45 complies with Russian
regulatory requirements and IAEA guidelines. The primary cooling mechanism for thereactor 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 isdesigned for a long service life.

93

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

NIKIET

Country of origin

Russian Federation

Reactor type

Boiling Water Reactor

Electrical capacity (MW(e))

45 – 50

Thermal capacity (MW(th))

180

Heat supply, base mode (Gcal/h)45 – 50

Expected capacity factor (%)

92

Continuous operation period (hr)19200

Design life (years)

80

2Plant footprint (m )9000

Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure(MPa)7.0oCore inlet/exit temperatures ( C)

180 / 286Core flow rate (ton/hr)

2600

Steam generating capacity (ton/hr)

320oSteam temperature ( C)

286

Pressure (MPa abs)

7.0

Main reactivity control mechanism

Control Rod Drive Mechanism (CRDM)

RPV height (m)11.15

RPV diameter (m)3.10

Module weight (metric ton)

250

Configuration of reactor coolant system

BWR

Power conversion process

Direct Rankine cycle

Fuel type/assembly array

UO2 pellet/hexagonal

Fuel assembly active length (m)1.98

Number of fuel assemblies

109

Fuel enrichment (%)

4.5

Fuel burnup (GWd/ton)

45.9

Fuel cycle (years)

7

Cogeneration capability

Yes

Approach to engineered safety systems

Passive

Number of safety trains

2

Refuelling outage (days)

30 (once in three years)

Distinguishing features

Designed for extreme Arctic and Northern area conditions

Modules per plant

1-4

Target construction duration (months)

48

Seismic design (g)3gCore damage frequency (per reactor-year)< 1E-6

Design Status

Conceptual design94

Nuclear Steam Supply System
The reactor employs a single-circuit heat removal system. The reactor steam removal system
is designed to transport 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 insome operational events when this system and 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. KARAT45 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 as shown in Figure 3 is located in the lower part of the reactor and consistsof 109 fuel assemblies (FA). The core has five complete FA rows and an incomplete sixthrow. 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 installed at
the core inlet for the coolant flow hydraulic profiling.

Fuel Assembly
Reflector

Core barrel

Reactor vessel

Figure 3: KARAT-45 reactor core (Reproduced courtesy of NIKIET)
Reactivity Control
The core includes 109 Control and Protection System (CPS) rods fitted in the FA guidechannels. 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.
95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 oflouver-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 - 11150 mm. The functions of the reactor
internals are:
•
•
•
•
•

positioning and retention of the reactor core;arrangement of coolant circulation;
steam separation;
heat transfer from the core;
vessel protection against neutron irradiation.

Reactor Coolant System
The reactor has a single-circuit heat removal system. The coolant circulation is natural. The
coolant (light water) flows through the reactor core upwardly while being heated and boiled.
The steam generated after drying is fed to the turbine before being discharged into thecondenser downstream of the turbine. Some part of the steam could be extracted from theturbine 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, thedeaerated condensate is fed into the reactor through feedwater supply headers. Inside thereactor, 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 the turbine 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, thesteam 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, thedefence-in-depth approach and a system of barriers to the release of radioactive materials into
the environment. The concept is aimed at preventing accidents and mitigating theirconsequences, should these occur. To achieve this, normal operation systems and safetysystems are required to perform three major functions:•••reactivity control;core cooling;
confinement 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 design-basis initiating event and during failure of any active or passive
component with mechanical parts independently on the initiating event (single failure96

principle). The safety system design also meets the requirement 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 thefeedwater supply and steam discharge systems failure. The system ensures the nuclear fuelcooling function. The system is based on a passive principle of action with heat removed
from the reactor through natural circulation.
Boron solution injection system
The boron solution injection system is designed to maintain the reactor subcritical while thepower system and independent power supply are disconnected, including accidents involving
coolant leakage. The criterion for the performance by the system of its function is that thereactor is maintained subcritical even in the event of CPS actuators failure and the reactor iscooled by water in pre-emergencies involving coolant leakage.
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 to organize the coolant movement. The emergency mitigation of
the primary coolant loss is ensured passively by draining water from the emergency
cooldown tanks into the reactor due to the gravitation because of the difference in the tankand reactor elevations.
Containment system
KARAT-45 reactor is located inside a reinforced concrete containment with a stainless steellining. The containment serves to localize accidents and is designed to withstand a pressureof up to 3 MPa. It forms an additional barrier to the leakage of radioactive materials into theenvironment while limiting, by its volume, the coolant loss during a reactor vessel break.
There are isolation gate valves installed on pipelines at the containment 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 and accidents. 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 and anticipated 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 coreheat by merely filling the core with coolant;
a substantial amount of coolant above the reactor core to ensure the reliable fuelcooling in majority of possible emergencies;
self-limitation of the in-vessel pressure variation rate due to the damping properties ofthe steam blanket.

97

7. Instrumentation and Control systems
In-core monitoring system is designed for monitoring of thermal-hydraulic and neutron
properties of the 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, thereactor water level, and temperature of water at FA inlet, in steam space and at thereactor 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 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 air-crash resistance buildings. The overall weight and size
parameters of the reactor unit due to its modular nature and transportability allow the deliveryof unit assembled at factory directly to the construction site by railway or other means oftransportation.
Reactor building
The reactor unit building performs the function of a primary containment. The reactor unit
accommodates KARAT-45 reactor, as well as systems responsible for emergency heatremoval from the reactor, emergency reactor shutdown and removal of radiolysis products
from beneath the reactor head. Besides, the reactor building includes an irradiated FA storage
facility and its cooling system, as well as the reactor facility’s moving and handlingequipment.
Control building
The control building houses the main control room and the emergency control room, electric
switchboard room, personnel rooms, communications center and other administrative rooms.
Balance of the plant:
Turbine Generator building
The turbine block accommodates the turbine generator, steam condensate line
components and equipment, and a bridge crane for moving operations. The dimensionsof the turbine block are 35x22 m, and its height is 28 m.
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
KARAT-45 design was developed in conformity with Russian laws, norms and rules for
NPPs and safety principles developed by the world community and IAEA recommendations.

98

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

Figure 1: Reactor System Configuration of KARAT-100(Reproduced courtesy of NIKIET)
2. Target Application
The KARAT-100 is a multipurpose BWR assigned for electricity generation, district heating
and for cogeneration. The KARAT-100 power unit has a high load following capability tocope with daily power variation from 20% to 100% of nominal capacity.

99

Figure 2: The KARAT-100 multipurpose application(Reproduced courtesy of NIKIET)
3. Development Milestones
2014

Development of Technical Proposal

2016

Conceptual design

2017-2019

Basic design, R&D

2020

Start of detailed design and equipment fabrication

2023

Commissioning

4. General Design Description
Design Philosophy
KARAT-100 reactor is being built as the base reactor for the evolution of power generationin isolated or remote locations not connected to the unified grid. The key factor that makes
this reactor a perfect choice for a nuclear cogeneration plant is its economic competitivenessagainst other sources of thermal and electric power, achieved primarily due to a combined
generation of heat (for district heating) and electricity.
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 thereactor directly to the turbine plant. The steam pressure in the system is 7 MPa, at the steam
temperature of 286 C. The humidity of the steam fed to the turbine plant is 0.1%. KARAT100 uses gravity-type steam separation with additional drying in louver-type separators
without centrifugal axial separators, which improves the stability of the reactor operation
100Reactor Core
The reactor core as shown in Figure 3, 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 thereflector). There are two types of FAs used in the reactor core: 111 with channels for the CPS
rods and 88 without such channels.
FA

FA with a control
rod
Reflector
BarrelReactor vessel

Figure 3: KARAT-100 reactor core
(Reproduced courtesy of NIKIET)
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 guidechannels and are grouped into clusters of 3 assemblies each to reduce the number ofactuators.
Reactor Pressure Vessel and Internals
KARAT-100 reactor vessel consists of a number of shells, a head and a bottom welded one toanother. The reactor internals - a barrel, a core support grid, a chimney, emergency heatexchangers 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 heatexchangers. All nozzles are located in the vessel’s upper part which guarantees that thenecessary volume of coolant is maintained even in the event of a nozzle break.

101

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer
Country of origin

NIKIET, Research and Development Institute of PowerEngineering
Russian Federation

Reactor type

Boiling Water Reactor

Electrical capacity (MW(e))

100Thermal capacity (MW(th))

360

Expected capacity factor (%)

92

Design life (years)

80

2Plant footprint (m )22500

Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure (MPa)7.0oCore inlet/exit temperatures ( C)

104 / 286

Main reactivity control mechanism

Control Rod Drive Mechanism (CRDM)

RPV height (m)13.25

RPV diameter (m)4.00

RPV or module weight (metric ton)

350

Configuration of reactor coolant system

Typical natural circulation BWRPower conversion process

Direct Rankine Cycle

Fuel type/assembly array

UO2 pellet/hexagonal

Fuel assembly active length (m)1.98

Number of fuel assemblies

199

Fuel enrichment (%)

4

Fuel burnup (MWd/kg U)

45.9

Fuel cycle (years)

7.5

Cogeneration capability

Yes

Approach to engineered safety systems

Passive

Number of safety trains

2

Refuelling outage (days)

30 (once in three years)

Distinguishing features
Modules per plant

Multi-purpose reactor: electricity generation, heat production andnuclear cogeneration plant
1-4

Target construction duration (months)

48

Seismic design

3g

Core damage frequency (per reactor year)

1E–6

Design Status

Conceptual design

Reactor Coolant System
The coolant is desalinated light water. The heat removal system is single-circuit. Steam isgenerated directly in the reactor vessel and, after being dried in the steam space, is fed to theturbine with a humidity of not more than 0.1 wt.% using gravity separation and louver-type
separators built in the reactor. The steam is dumped into the condenser downstream of theturbine. Some part of the steam could be extracted for heating of plant system’s in housewater and for heat supply.
102

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 radiologicalimpacts on the personnel, the public and the environment during normal operation and incases of operational occurrences and emergency events. KARAT-100 safety is ensured
through the technological sophistication of design, the required 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 qualification and discipline of personnel.
Engineered Safety System Approach and Configuration
KARAT-100’s system of engineered and organizational measures forms five defence-indepth 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 the help of coil-type emergency heat exchangers accommodated inside thereactor vessel and emergency cooling tanks. The system is based on a passive principle ofaction. 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:
•
•
•
•
•
•

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 safetysystems:••

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 theborated coolant back from the reactor cavity in the event of a reactor vessel or nozzle
break;

103••The reactor water accumulation system aimed at keeping the water inventory in theaccumulators for making up the reactor in emergencies caused by a decrease in thereactor vessel coolant level;
The emergency power supply system used in the event of a loss of power supply fromthe 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 eventof reactor vessel break. There are isolation gate valves installed on pipelines at thecontainment 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 incases of operational occurrences and emergency events. KARAT-100 safety is ensured
through the specific transfer and distribution of radioactive substances due to water boiling.
The key factors are:••A high inter-phase barrier (water-steam) to the spreading of nongaseous radionuclides
prevents these from entering the steam-condensate line;
Continuous degassing of coolant and removal of gaseous fission products from thecircuit limit their accumulation in the circuit.

7. Instrumentation and Control systems
In-core monitoring allows thermal-hydraulic and neutronic parameters of the reactor core andthe in-core coolant natural circulation circuit to be measured directly and indirectly in
different operating modes of the reactor. KARAT-100 reactor is expected to be monitored forneutronic and thermal parameters, including the reactor water level, the core inlet watertemperature, and the steam space temperature, and periodically tested for the water chemical
properties.
8. Plant Arrangement
The building layout plan for the land-based power unit of KARAT-100 is designed in such a
way that the reactor system, including its servicing systems, spent fuel pool, and auxiliaries
are located in double protective air-crash resistance buildings. The designers also claim thatthe overall size of the steam generating unit allows transportation of the reactor by railway.
Reactor building
The reactor unit building performs the function of a primary containment. The reactor unit
houses the KARAT-100 reactor as well as the systems responsible for the emergency removalof heat from the reactor, the emergency reactor shutdown and the removal of radiolysis
products from beneath the reactor head. Besides, the reactor building houses an irradiated FA
storage facility and its cooling system, as well as the reactor facility’s handling equipment.

104

Figure 4: Plant Layout arrangement of the KARAT-100(Reproduced courtesy of NIKIET)
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:
Turbine Generator building
The turbine block houses the turbine generator, the steam condensate line components
and equipment, and a bridge crane for moving operations. The dimensions of theturbine block are 42x28 m, and its height is 28.4 m.
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.

105

ELENA (Kurchatov Institute, RussianFederation)
1. Introduction
The ELENA nuclear thermoelectric plant (NTEP) is a direct conversion water-cooled reactor
without on-site refuelling capable to supply 68 kW(e) of electricity and 3.3 MW(th) of
heating capacity for 25 year without refuelling. The technology and techniques weredeveloped incorporating the experience from the construction and operation 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 has been developed by the Russian ResearchCentre “Kurchatov Institute” (RRC KI). The ELENA NTEP is a land-based plant; however,in principle it is also possible to develop the underground or underwater versions. The reactor
and its main systems are assembled from factory-fabricated finished components or modules,
whose weight and dimensions enable any transport delivery 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 and other external reactivity disturbances
without moving the control rods;
The use of thermoelectric energy conversion to produce electricity.

Figure 1: Reactor system configuration of ELENA(Reproduced courtesy of the Kurchatov Institute)106

2. Target Application
The unattended ELENA NTEP plant is designed to produce heat for towns with a populationof 1500–2000 located in remote areas where district heating is required. Since it is auxiliaryin nature, the electricity generation of 68 kW could be used for the in-house power needs ofthe plant and to supply electricity to consumers requiring a highly reliable power supply, suchas hospitals, etc. A desalination unit can be used in combination with the ELENA NTEP.
3.

General Design Description

Design Philosophy
The ELENA reactor is designed with an integrated primary circuit. The design features of
ELENA ensure high reliability and safety, eliminate adverse environmental impacts, andmake the ELENA NPP an attractive source of heat and power supply for small settlements
located in remote areas, including seismic and draught ones, as well as in uninhabited orunderwater stations, e.g., robotized systems for investigation and extraction of ocean
resources or hydrology research laboratories.
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) consists of a reactor core and internals and thesteam generators. The design is based on an integral reactor located in the large volume ofsecondary water. The NSSS is enclosed in a cylindrical vessel that is embedded in the reactor
pool structure which is filled with water. Electric power is generated in semiconductor
thermal battery due to the temperature difference provided between primary and secondary
circuits.
Reactor Core
Pellet type uranium dioxide fuel is used with the average U235 enrichment of 15.2%; theneutron moderator and coolant is water specially treated according to specified waterchemistry. Cylindrical fuel elements with stainless steel cladding are installed in 109 fuelassemblies of 55 fuel elements each; 216 absorber rods with boron carbide based neutron
absorber are divided into 6 groups. Fuel assemblies also include burnable absorbers made of
Gd-Nb-Zr alloy. The U235 load is 147 kg.
Reactivity Control
A reliable operation and reactivity control is achieved through the implementation of passive
reactivity regulation and control systems. The control and safety systems, including thecontrol rods and control rod drive mechanisms are used for reactivity control. The control andsafety systems are designed to be fail safe. The ELENA reactor target to provide the total
reactivity margin in a hot core as small as to secure the survival of an 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 the equivalent diameter of 833 mm isinstalled in a steel shell with a diameter of 920 mm and is encircled by an iron-water shield.
The strengthened stainless steel reactor vessel has an internal diameter of 1250 mm and aheight of 3700 mm with a wall thickness of 132 mm.

107

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

Russian Research Centre “Kurchatov Institute” (RRC KI)

Country of origin

Russian Federation

Reactor type

PWR

Electrical capacity (kW(e))

68Thermal capacity (kW(th))

3300

Expected Capacity Factor> 87

Design Life (years)–

2

Plant Footprint (m )–Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure (MPa)19.6oCore inlet/exit temperatures ( C)311/328

Main Reactivity Control Mechanism

Control rod drive and absorber rods

RPV Height (m)3.7

RPV Diameter (m)1.25

Weight (without primary and secondary coolant) (ton)160

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Heating reactor, direct, thermo-electric

Fuel Type/Assembly Array

UO2 pellets; MOX fuel is an option

Fuel Assembly Active Length (m)–Number of Fuel Assemblies

109

Fuel Enrichment (%)

15.2

Fuel Burnup (GWd/ton)

57600/27390

Fuel Cycle (months)

300

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains–

Refueling Outage (days)–

Distinguishing features

Modules per Plant

Twenty five (25) years life of the plant without refuelling, passive
reactivity regulation and control systems and a very long period of
unattended operation.
1

Target Construction Duration (months):

48 (including manufacturing of components)

Seismic design–

Predicted core damage frequency (per reactor year)–

Design Status

Conceptual design

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 unattended operation without on-siterefuelling for up to 25 years. The temperature of the water within the third loop is ~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 enclosed108

in the draft tube for excess heat discharge to the atmosphere.
Primary Cooling Mechanism
The ELENA has three water coolant loops. The primary coolant loop is completely contained
within the secondary 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 apressure 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 heatexchanger of circuits II–III; the coolant is specially treated water, which also acts aspart of the steel-water radiation shielding;
Circuit III is designed as a thermo-siphon with water or low-boiling coolant. Circuit
III transfers heat through 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.

4. Safety Features
The reactor is installed in a caisson forming a heat-insulating gas cavity in the area of thestrengthened reactor vessel and a caisson space above the reactor cover to house control andprotection system (CPS) drives and to prevent radioactive substances from escaping into thesurrounding space in case of a circuit I break. The localizing safety systems provide thedefence in depth and secure the plant safety based on inherent safety features andpredominantly passive phenomena; they require no human intervention or external powersources. The safety barriers of the ELENA-NTEP are:
•
•
•
•
•

The fuel elements;
The leak-tight primary circuit;
The caisson;
The reactor vessel and the guard vessel designed to withstand the pressure arising
within each of them at their consecutive failure; andAn embedded silo sealed with the protective plate.

Special measures for the protection of hot water consumers ensure that radioactivity is neverreleased 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 defence-in-depth approach based on
six safety barriers prevents the primary circuit from depressurization and secure activity
confinement inside the reactor during accidents. The use of a self-adjustable water-cooled
reactor coupled with thermoelectric mode of heat conversion and natural circulation of
coolant makes it possible to exclude movable elements from the technological circuit of an
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 theconsumers. The active components of the protection system are scram actuators for six
compensation groups of the control rods.

109Decay heat removal system
A low specific thermal power of the ELENA reactor enables easy removal of the residual heat
after the reactor shutdown. The residual heat is damped naturally to the compartment and thefuel elements are not super-heated during 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; italso periodically provides information on the state of the plant.
Containment system
The reactor is installed in a caisson forming a heat-insulating gas cavity in the area of thestrengthened reactor vessel and a caisson space above the reactor cover to house control andprotection system (CPS) drives, and to prevent radioactive substances from escaping into thesurrounding 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 of radioactivity; water that fills thecontainment volume is the 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.
5. Plant safety and Operational Performances
The ELENA reactor does not require an operator during nominal power operation of theplant. Operators are required for assembly, startup and to begin nominal operation. The
reactor is designed to operate in a base load mode. The reactor installation is based on passive
principles of heat removal (natural convection in all circuits, except for heat transport to theconsumers) in normal operation and in shutdown conditions. A decrease in the heat orconsumer 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 mechanical parts 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 forthe 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 andpressure in the coolant loop is monitored, but do not provide feedback through the control
loop during startup. To begin the operation, the poison rods are pulled completely from thecore, and are never inserted during nominal operation. To start up and reliably shut down thereactor 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 an individual drive.
6. Instrumentation and Control systems
The instrumentation and control (I&C) system of the ELENA reactor is specially arrangement
to register parameter deviations at early stages of the accidental conditions to predict theirfurther progression.
7. Plant Arrangement
The reactor system can be broken into two parts for shipment. It is possible to fuel the system110

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; anauxiliary power supply system; and a radiation monitoring system, including process
radiation monitoring, dosimetric monitoring, and environmental monitoring.
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 aremonolithic reinforced concrete. The vessel head of the system is removable. The plant
incorporates a physical protection system, has a fence and is equipped with external lighting.

Figure 2: Plant general view of ELENA(Reproduced courtesy of the Kurchatov Institute)
Control building
The plant has a main control and monitoring room accommodating the start-up andinstrumentation and control equipment, as well as the equipment necessary to prepare theinformation to be transmitted to a monitoring centre.
Balance of Plant:
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 thereactor nominal operating mode simultaneously with heat transfer to circuit II. Thispower is used for the plant auxiliary needs; it could also be supplied to a small town
without district power supply, partially replacing a diesel power plant. The TEG
111

consists of eight identical thermoelectric units (TEU). Each of them includes 36
thermoelectric modules equipped with thermoelectric packs of bismuth tellurides withelectronic and hole conduction.
Electric power systems
The ELENA-NTEP CSS has three independent power supply system: i.e., two (2) TEG
sections, a diesel generator, and a storage battery. The electric power output can becontrolled either by the use of shut resisters or by short circuiting the TEs. The TE
power conversion system has low electrical conversion efficiency, and the waste heat isused for district heating.
8. Design and Licensing Status
The assembly drawings of the ELENA have been completed, and are ready for fabricationand testing of the system.
9. Plant Economics
The target cost of electricity in US $ is 0.06/kW-hour. The cost of energy for district heating
is expected to be US$ 50/GCal. The ELENA NTEP is designed for long-term operation witha single fuel load and absence of the personnel, thus the labour payment and life support costs
reduce considerably. Although the specific construction cost of the ELENA NTEP is
projected to be higher, the power generation cost is expected to be competitive with theorganic fuel for remote areas or for specific applications.

112

RUTA-70 (NIKIET, Russian Federation)
1. Introduction
RUTA-70 is a water-cooled water-moderated integral pool-type heating reactor with athermal capacity of 70 MW(th). RUTA-70 is a dedicated reactor for district heating or potablewater production; it has no energy conversion system. The reactor pool is made of reinforced
concrete lined by stainless steel. In the primary cooling circuit, the heat from the core istransferred to the primary heat exchanger (HX) by forced convection at full power but by
natural convection for powers 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
make it possible to increase the coolant flow rate in the primary circuit and raise the downcomer temperature by reducing the water thermal gradient in the reactor core. The reactor
power is regulated using the control and protection system (CPS) drives, which are installedin the reactor pool water. The period of continuous operation of the reactor equipment
without any need of maintenance is about one year. Simplicity, high reliability and inherent
safety features of the RUTA are based on the low pressure and temperature of the primarycoolant as well as integral design of the reactor. Due to safety features, nuclear district
heating plant (NDHP) using RUTA reactors can be located in the immediate vicinity of theheat users. Specific features of the RUTA are as follows:
•
•
•
•
•

No high coolant pressure in the primary circuit (atmospheric pressure above the waterlevel in the reactor pool);
Large water inventory in reactor pool enables heat accumulating capacity;
Low core power density (< 30 MW(th)/m3);
Natural convection is applied in core cooling for lower level below 30% of the rated
power, as well as for cooldown modes;
A three-circuit system of heat supply to consumers with the lowest water pressure
kept in the primary circuit (primary pressure < secondary pressure < consumer circuit
pressure).

2. Target Application
The conceptual design of RUTA is primarily developed to provide district heating in remotely
isolated areas of the Russian Federation. Continuous increase of the organic fuel costs in the
country essentially enhances the prospect of RUTA as a heating reactor. In addition, anotherpotential thermal energy application of the RUTA is seawater and brackish water
desalinations based on distillation process.
3. Design Milestones
1990

Conceptual design of the 20 MW(th) RUTA heating plant

1992

Feasibility study ‘Design and Development of the underground nuclear heating plant with the RUTA reactor for
district heating in Apatity-city, Murmansk region’

1994

Feasibility study ‘Underground NHP with 4 × 55 MW(th) RUTA reactors for district heating in Apatity-city,
Murmansk region’

2003

Technical and economic assessments for using the 70 MW(th) RUTA reactor to upgrade the district heating
system in Obninsk, Kaluga region.
Feasibility study of constructing a NDHP in the State Research Centre of the Russian Federation - IPPE,
Obninsk. Approval of the project by the Board for the programme ‘Development of Obninsk as a science town’113Figure 1: Reactor System Configuration of RUTA-70 (Reproduced courtesy of NIKIET)
4. General Design Description
Design Philosophy
The basic design principles of this reactor are simplicity of the design and a high safety level
due to a low pressure and a large coolant inventory in the primary system. The concept of
RUTA-70 reactor facility is based on simple design and, as a result, a low cost of plant
construction and operation, and high level of safety achieved through the specific features ofdesign and inherent safety mechanisms. The reactor facility is a part of the ground based
nuclear heating plant (NHP) designed similarly to research reactors of the pool type.
Nuclear steam supply system
The RUTA reactor facility has a two-circuit layout. The primary circuit is an in-pile reactor
core cooling circuit and the secondary circuit is an intermediate one that removes heat fromthe reactor and transfers it to the third circuit, which is the consumer circuit, i.e., to theheating network. Heat is transferred from the primary circuit to the secondary circuit andfrom the secondary circuit to the third circuit through the leak-tight heat-exchange surfaces,
which are used to avoid the spread of radioactive products from the reactor circuit to theconsumer. Most of the plant equipment, including the primary-to-secondary side heatexchangers (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 thechimney section. The core is designed with the ‘Cermet’ fuel rods that contribute to thereactor safety due to a high thermal conductivity of the fuel matrix and its role as the
additional barrier to the fission products release.

114

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

NIKIET

Country of origin

Russian Federation

Reactor type

Pool-type

Electrical capacity (MW(e))

N/AThermal capacity (MW(th))

70

Expected capacity factor (%)

95

Design life (years)

60

2Plant footprint (m )100000

Coolant/moderator

Light Water

Primary circulation

Natural (below 30% of N)/Forced (for 30-100% of N)

System pressure (MPa)Atmospheric pressure at reactor poll water surfaceoCore inlet/exit temperatures ( C)

75 / 102

Main reactivity control mechanism

Control rod driving mechanisms, Burnable poison

Reactor pool height (m)17.250Lower part of reactor pool diameter (m)3.2

RPV weight (ton)-

Configuration of reactor coolant system

Integrated pool-type

Power conversion process

N/A

Fuel type/assembly array

Cermet (0.6 UO2+ 0.4 Al alloy) / hexagonal

Fuel assembly active length (m)1.4

Number of fuel assemblies

91

Fuel enrichment (%)

3.0

Fuel burnup (GWd/ton)

25–30

Fuel cycle (months)

36

Cogeneration capability

Yes, for desalinations using distillation

Approach to engineered safety systems

Passive

Number of safety trains

2.0

Refuelling outage (days)

28

Distinguishing features

Designed for low temperature process heat, coupling with desalination
system, radioisotopes production for medical and industrial purposesand creation of neutron beams for neutron therapy

Modules per plant

1.0

Target construction duration (months)

36

Seismic design (g)>0.8 (automatic shutdown)

Predicted core damage frequency (per reactor year)

PSA is not completed, assessment < 1E–6

Design status

Conceptual design

The reactor core consists of 91 fuel assemblies (FA) of hexagonal geometry with 120 fuelrods per each FA. The height of reactor core is 1400 mm or 1530 mm depending on the fuelrod 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 fuel assembly with the fuelassembly pitch of 147 mm and the fuel rod pitch of 12.2 mm. In the core, there are 42 reactor
control and protection system (RCP) rods composing two shutdown systems with diverse
actuators.

115

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 andmovable control rods. The reactivity control is performed by regulating the control rods andutilizing 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. The
control and protection system (CPS) controls the reactor power using the drives submerged
into the reactor pool.
Reactor Coolant System
The primary coolant forced circulation is provided by two main circulation pumps - onepump per each of two reactor loops. Two MCPs of axial type are installed in the bypass linesof the main circulation loop close to the down-comer inlet. Pumps are located to allow easy
access for inspection or replacement. The loop arrangement of the primary circuitcomponents, with the secondary circuit pressure exceeding the pool water pressure, ensuresthat the reactor coolant is localised within the reactor tank.
Primary Cooling Mechanism
The heat from the core to the primary heat exchanger (HX) is transferred by forced
convection of the primary water 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 make itpossible to increase 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 the upper part of the shell of the chimney section. Pipelines of watersupply to the primary HXs are connected to the header from both sides. HXs are located in
the upper part of the pool. Under the natural convection operating mode, the forced
convection circuit is isolated by a check valve installed at the pipeline of the individual
section. Using the same valve, the natural convection circuit can be isolated. 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 from the pumphead 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 distinctive feature of pool type reactors is that there is no excess pressure in the reactor
pool; this excludes an accident with an instantaneous rupture of the primary circuit andcessation of heat transfer from the core due to dry out. The high safety level of pool reactors
is achieved through their design features, which make it possible to resolve some of the major
safety issues through the employment of the naturally inherent properties of the reactor. 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 onprinciples such as multi-channelling, redundancy, spatial and functional independence,
application of a single failure criterion and diversity. The acceptable safety level isunderstood as the level when the total effective annual internal and external exposureradiation dose for the population under normal operating and emergency conditions does notexceed the natural background dose.

116

Engineered Safety System Approach and Configuration
The RUTA-70 uses mostly passive systems to perform safety functions such as: air heat sinksystem for emergency cooldown (ASEC), gravity driven insertion of the control rods in thecore as reactor safety control system, the secondary circuit overpressure protection system,
the overpressure protection system for air space in the reactor pool and pre-stressed concrete
external impacts protection system. The computational analysis of beyond design basis
accidents has revealed that in case of multiple failures in the reactivity control systems anddevices, safety can be ensured by self-control of reactor power (boiling - 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. This means that the heating upof the core structural components, including fuel, or the boiling of water in the core results inspontaneous reduction or self-limitation of the reactor’s nuclear power at any initial positionof the control rods, including the scram rods. The HXs can be passively flooded by waterfrom the pool providing long term heat sink in case of emergency. Pumps and HX are locatedas such to allow easy access for inspection or replacement. The protective flooring composedof slabs that are installed above the reactor pool to avoid possible damage to the primary
components from external impacts. To prevent gas and vapour penetration to the reactor hallfrom the upper part of the reactor, joints of the protective slabs are gas-tight. An
unrecoverable leak in the reactor pool will not occur due to the concrete pool which isdesigned to withstand external events, including maximum design basis earthquake and water
filtration to the ground in a beyond design basis accident.
Decay Heat Removal System
Natural circulation in the secondary circuit provides for residual heat removal from the
shutdown reactor and passive cooldown of the reactor facility in blackout emergency
situation. The passively actuated ASEC provides residual heat removal to the ultimate heatsink (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 isconnected at the bypass line of the network heat exchangers. If all controlled trains of heatremoval are lost, heat losses via the external surface of the reactor pool to the surrounding
environment (ground) are considered as an additional safety train. Residual heat isaccumulated in the pool water. The transient of pool water heat-up in the aqueous mode
before the onset of boiling takes several days. As soon as boiling starts, steam goes to thereactor hall where it is condensed (passive condensing facilities are provided). A reactor boiloff without makeup takes 18 to 20 days. Upon completion of this period residual heat isbalanced by heat transfer to the ground. Core dry out is avoided. Moderate temperatures are
not exceeding the design limits characterize fuel elements.
Emergency Core Cooling System
In emergency situations, residual heat is transferred by natural circulation of the coolant in
the reactor tank and in 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 anactive system.
Reactor pool
The reactor pool consists of reactor core and internals, control and protection system,
distributing and collecting headers and a large amount of water. Water in the reactor pool
have a high heat accumulating capability that ensures slow changing of coolant parameters

117during transient and emergency conditions 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 both in the base load and load follow modes. Two
independent systems based on 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 thosein 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 this system can be inserted in the core under the effect of gravity.
7. Plant Arrangement
Reactor building
The protective flooring composed of slabs is installed above the reactor pool to avoid
possible damage to the primary components from external impacts. To prevent gas andvapour penetration to the reactor hall from the upper part of the reactor, joints of theprotective slabs are gas-tight.

2

10
9

5

8

4

6

1

3

7

Figure 2: RUTA-70 Plant layout (Reproduced courtesy of NIKIET)(1) Core; (2) Primary heat exchanger; (3) Check valve; (4) Pump;(5) Primary circuit distributing header; (6) Primary circuit collecting header;(7) Secondary circuit inlet pipeline; (8) Secondary circuit outlet pipeline;(9) SCS drives; (10) Upper slab

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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 to electrical devices and systems, instrumentation and control.
The respective services are responsible for radiation monitoring during operation and repair.
A supervising physician and a refuelling operator are added to the regular shift staff for thecore fuelling, first core critical mass attaining, power startup and refuelling periods. With
managers, engineers and technicians working in one shift (a chemical engineer, an engineer
for electrical devices and systems, an engineer for instrumentation and industrial control) andthe administrative staff taken into account, the total personnel number is about 40 persons.
Balance of Plant:
The turbine and associated systems are not used in the NHP RUTA-70. RUTA-70 is a multipurpose NHP for district heating, desalination and radioisotopes production for medical andindustrial purposes. The principal scheme of NHP with RUTA reactor integration into the
district heating system along with peak/backup boiler is show in Figure 3 below. Such
scheme in spite of some reduction of autonomy of district heating system (in comparison tohigh-temperature reactors) possesses several major advantages: 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 due to system
optimization.

Figure 3: Scheme of NHP with RUTA reactor integrated in district heating system(Reproduced courtesy of NIKIET)(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.

119

8. Design and Licensing Status
To provide an operating reference for the reactor, the Institute of Physics and PowerEngineering (IPPE) proposed to build the first reference RUTA-70 unit in Obninsk, where theinstitute is located. The proposal, developed in 2004, was supported by the feasibility study
jointly carried out by NIKIET, IPPE, and Atomenergoproekt (Moscow). This study showedthat RUTA-70 should be ultimately deployed along with the non-nuclear sources of power
operating in peak and off-peak mode. For the current status, RUTA is still in the conceptual
design stages.
9. Plant Economics
The results of economic assessments show vast potential market for nuclear heating plantwith RUTA-70 reactors. Duration of the NHP construction period is 3 years, including
adaptation of the standard design to local site conditions. The projected specific capital costin US $ is 279.6 per kW(th). Simplicity of design, low parameters, space-saving heatexchanging equipment and the use of passive systems contribute to the reduction of capitalcosts for construction of the NHP RUTA-70. The annual operating costs for the NHP RUTA70 is 1772.5 US $ per year. The O&M costs can be reduced due to the limited number ofrequired operating personnel and low costs for the repair and consumables. Further R&D is
being performed for the reduction of operating costs by changeover to operating life of 100
years.

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NuScale (NuScale Power Inc.,
United States of America)
1. Introduction
The NuScale Power Module (NPM) is a small, light water cooled pressurized-water reactor(PWR).The NuScale plant is scalable and can be built to accommodate a varying number of
NPMs to meet a customer’s energy demands. The 50 MW(e) NPM provides power inincrements that can be scaled to 600 MW(e) gross in a single facility. A twelve module
configuration is the current reference plant size for design and licensing activities. Each NPM
is a self-contained module that operates independently of the other modules in a multimodule configuration, as shown in Figure 1. All modules are managed from a single control
room. The plant design contains following significant features:
•
•
•
•
•
•
•

reliable and passively safe systems that are simple in design and operation;
safety features that assure a core damage frequency substantially lower than thecurrent light water reactor fleet;reactor building designed to protect against aircraft impact while maintaining
spent fuel pool integrity and preserving containment integrity;
60-year design plant life;
reduced construction schedule compared to nuclear plants of a comparable output;
modularization to enable in-shop fabrication of reactor and containment
components; anduse of half-height traditional 17x17 nuclear fuel assemblies.

Figure 1: Cut-away view of NuScale power plant(Reproduced courtesy of NuScale Power Inc.)
2. Target Applications
NuScale design is a modular reactor for electricity production and non-electrical process heatapplications.

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3. Development Milestones
2003

First Integral test facility operational

2007

NuScale Power was formed

2008

US NRC design certification pre-application started

2012

Twelve-reactor simulated control room was commissionednd

NuScale to submit design certification application to the NRC

ndFirst customer to submit application for a combined construction and operation licence(COL)
NuScale design certification target

2016 (2 half)2017 (2 half)2020

4. General Design Description
Design Philosophy
The NuScale plant design philosophy consists of: a design that uses proven light-water
reactor technology, modular nuclear steam supply system, factory-fabricated power modules
with shippable components, and passive safety systems that allow for unlimited coping timeafter a design basis accident without power, operator action, or makeup water. The NPM isdesigned to operate efficiently at full-power conditions using natural circulation as the meansof providing core coolant flow, eliminating the need for reactor coolant pumps.
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) consists of a reactor core, helical coil steam
generators, and a pressurizer within a reactor pressure vessel (RPV). The NSSS is enclosed in
an approximately cylindrical containment vessel (CNV) that sits in the reactor pool structure
as shown in Figure 2. The reactor core is located below the helical coil steam generators
inside the RPV as shown in Figure 3.

Figure 2: Nuclear Steam Supply System

Figure 3: Diagram of the NuScale Power
Module(Reproduced courtesy of NuScale Power Inc.)

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MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

NuScale Power, LLC

Country of origin

USA

Reactor type

Integral PWR

Electrical capacity (MW(e))

50 (gross)

Thermal capacity (MW(th))

160

Expected Capacity Factor

˃95%

Design Life (years)

60

2Plant Footprint (m )130000

Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure

12.8 MPaoCore inlet/exit temperatures ( C)-- / 302

Main Reactivity Control Mechanism

Control rod drive, boron

RPV Height (m)17.4

RPV Diameter (m)2.9Module Weight (metric ton)--

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine Cycle

Fuel Type/Assembly Array

UO2 pellet/17x17 square

Fuel Assembly Active Length (m)2Number of Fuel Assemblies

37

Fuel Enrichment<4.95%

Fuel Burnup (GWd/ton)

TBD

Fuel Cycle (months)

24

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

2

Refueling Outage (days)

10

Distinguishing features

Unlimited coping time for core cooling without AC or DC
power, water addition, or operator action as demonstrated

Modules per Plant

1–12

Target Construction Duration (months)

36

Seismic design

0.5g peak ground acceleration

Predicted core damage frequency (per reactor year)

1E-8 (internal events)

Design Status

Under development

Reactor Core

The core configuration for the NPM, as shown in Figure 4, consists of 37 fuel assemblies and
16 control rod assemblies. The fuel assembly design is modelled from a standard 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 of standard plant fuel and is
supported by five spacer grids. The fuel is UO2 with Gd2O3 as a burnable absorber123homogeneously mixed within the fuel for select rod locations. The U235 enrichment is below
the current U.S. manufacturer limit of 4.95 percent enrichment. The reactor has a refuelling
cycle of 24 months.

Figure 4: Reactor Core Assembly (Reproduced courtesy of NuScale Power Inc.)
Reactivity Control
Reactivity control in 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 symmetricallylocated in the core, functions as a regulating group that is used during normal plant operation
to control reactivity. The shutdown group comprising 12 rods is used during shutdown andscram events. Control rods absorber material is B4C and the control rod length is 2 meters.
Reactor Pressure Vessel and Internals
The RPV consists of an approximately cylindrical steel vessel with an inside diameter ofapproximately 9 ft. and an overall height of approximately 58 ft. and is designed for an
operating pressure of approximately 1850 psia. The upper and lower heads are torisphericaland the lower portion of the vessel has flanges to provide access for refuelling. 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 primary systempiping.
Reactor Coolant System
NPM has natural convection-cooled core which does not require reactor coolant pumps. The
NSSS and CNV are immersed in reactor pool which ensures passive long-term cooling andenhanced fission product retention. The reactor core is located inside a shroud connected tothe hot leg riser. The reactor core heats reactor coolant causing the coolant to flow upward
through the riser. When the heated reactor coolant exits the riser, it passes over the tubes of
the helical coil steam generators, which act as a heat sink. As the reactor coolant passes overthe steam generator tubes, it cools, increases in density, and naturally circulates down to thereactor core, where the cycle begins again.

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Steam Generator
Each NPM uses two 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. The steam generator consists of tubes connected to feed and steam
plenums with tube sheets. Preheated feedwater enters the lower feed plenum through nozzles
on the RPV. As feedwater flows through the interior of the steam generator tubes, heat is
added from the reactor coolant. The secondary side fluid is heated, boiled, and superheated to
produce steam for the turbine-generator unit.
Pressurizer
The 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 in the reactor coolant system is reduced using spray provided by thechemical and volume control system (CVCS).
5. Safety Features
Engineered Safety System Configuration and Approach
Each NPM incorporates several simple, redundant, and independent safety features, which
are discussed as follows:
Decay Heat Removal System
The decay heat removal system (DHRS) as shown in Figure 5, provides secondary side
reactor cooling for non-LOCA events when normal feedwater is not available. The system isa closed-loop, two-phase natural circulation cooling system. Two trains of decay heatremoval equipment are provided, one attached to each steam generator loop. Each train iscapable of removing 100 percent of the decay heat load and cooling the reactor coolant
system. Each train has a passive condenser immersed in the reactor pool. During normal
operations, the condensers are maintained with sufficient water inventory for stable andeffective operation.
Emergency Core Cooling System
Emergency core cooling system (ECCS) as shown in Figure 6, consists of two independent
reactor vent valves (RVVs) and two independent reactor recirculation valves (RRVs). For
LOCAs inside containment, the ECCS returns coolant from the CNV to the reactor vessel.
This ensures that the core remains covered and that decay heat is removed. The ECCS
provides a defense-in-depth means of decay heat removal in the unlikely event of a loss offeedwater flow, combined with the loss of both trains of the DHRS. The ECCS removes heatand limits containment pressure by steam condensation on, and convective heat transfer to,
the inside surface of the CNV. Heat is then transferred by conduction through the CNV walls
and convection to the water in the reactor pool. Long-term cooling is established via
recirculation of reactor coolant to the RPV through the ECCS recirculation valves.
Reactor Pool
The reactor pool is a large stainless steel lined pool located below the plant ground level.
Water in the pool provides cooling for the NPM for a minimum of 72 hours following any
design basis accident. During normal plant operations, heat is removed from the pool througha cooling system and ultimately rejected into the atmosphere through a cooling tower or otherexternal heat sink. In an accident where off-site power is lost, heat is removed from the
NPMs by allowing the pool to heat up and boil. Water inventory in the reactor pool is large

125enough to cool the NPMs for an unlimited time without adding water.

Figure 5: Decay Heat Removal System
Figure 6: Emergency Core Cooling System(Reproduced courtesy of NuScale Power Inc.)
Containment System
The major safety functions of the Containment Vessel (CNV) are to contain the release ofradioactivity following postulated accidents, protect the RPV and its contents 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 approximately 15 ft. and an
overall height of approximately 76 ft. 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 a passive heat sink for containment heat removal under
LOCA conditions. The CNV is designed to withstand the environment of the reactor pool aswell as the high pressure and temperature of any design-basis accident. The CNV pressure ismaintained at a vacuum under normal operating conditions. Maintaining a vacuum provides
the thermal insulation of the RPV to parasitic heat loss during normal operation. The vacuum
also reduces moisture inside containment that could contribute to component corrosion andimpact the reliability of instrumentation and other systems within the CNV. Following an
actuation of the ECCS, heat removal through the CNV rapidly reduces the containmentpressure and temperature and maintains them at less than design conditions indefinitely.
Steam is condensed on the inside surface of the CNV, and is passively cooled by conduction
through the CNV and convection to the reactor pool water.
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 is disassembled into to three major components: the lower RPV section that
contains the core and lower internals, the lower CNV section, and the upper RPV/CNV
section that contains the steam generators and pressurizer. After refuelling the core andinspecting the module sections, the module is reassembled and moved to its operations bay
and reconnected to steam and feedwater lines. Other modules in the plant continue to operate
while each module is refuelled.

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7. Instrumentation and Controls Systems
The NuScale instrumentation and controls systems provide automated safety responses tospecific initiating events such as reactor trip, followed by an integrated safety actuation of
one or more of the passive safety systems; the DHRS and ECCS. The control room
habitability system (CRHS) ensures that plant operators are adequately protected against theeffects of accidental releases of toxic and radioactive gases. Areas served by the CRHS are
maintained at positive pressure relative to adjacent areas. Compressed breathable air storage
capacity can provide clean air to the control room spaces for at least 72 hours following an
initiating event. The NuScale design effectively integrates human factors engineering (HFE)
into the development, design, and evaluation of the plant. HFE employs state-of-the-art
principles to ensure NuScale products facilitate safe, efficient, and reliable performance of
operations, maintenance, tests, inspections, and surveillance tasks. The NuScale design
minimizes human error through fail-safe design functionality, allows multi-modular control
capability with effective automation design, employs digital display design and soft control
technology to enhance usability, and provides optimum workload management.
8. Plant Arrangement
Reactor building
The NuScale plant consists of a power generation complex and common facilities. The power
generation complex for the NuScale plant consists of up to 12 power generation units (12NPMs and associated turbine generators), module assembly equipment, fuel handlingequipment, turbine maintenance equipment, and radioactive waste processing equipment. A
12-unit plant net total output is approximately 570 MW(e). Figure 7 presents the layout of amulti-unit plant. Each NPM is in its own bay immersed in the common reactor pool. Eachreactor bay is approximately 20 ft. wide by 20 ft long and built into the nearly 75 ft. deep
pool with a normal water level of approximately 69 ft. Each bay has a concrete cover thatserves as a biological shield. The cover also serves to prevent deposition of foreign materials
onto an NPM. The reactor pool is located in a Seismic Category I building designed towithstand postulated adverse natural conditions and aircraft impact.
reactor building cranecontainment vessel
flange tool

reactor vessel
flange tool

spent fuel pool

refueling machine

NuScale Power
Modulesreactor poolmodule importtrolley

Figure 7: Reactor building plan view (Reproduced courtesy of NuScale Power Inc.)
Control building
The main control room is housed below grade in the control building located adjacent to thereactor building. The plant will have a control console in the main control room for all
installed units. Each reactor operator will monitor and control multiple units from the control

127room console. The reactor operators monitor the automated control system for each reactor.
Each reactor is outfitted with monitors provided with soft controls and some select manual
push buttons for operator control. The supervisor station provides an overview of all reactors
using multiple monitors. All monitor displays are designed using human factors analysis toenhance simplicity. The display layout and design uses graphical representations of plantsystems and components.
Balance of Plant:
Turbine Generator Building
A NuScale plant has two separate turbine buildings, each housing up to six turbinegenerators. The turbine buildings are above-grade structures that house the turbinegenerators with their auxiliaries, the condensers, condensate systems, and the feedwater
systems. An emblematic turbine generator is shown in Figure 8. Each turbine-generator
is associated with a single NPM and has dedicated condensate and feedwater pumps.
Each turbine-generator is supported by an above-grade pedestal. Each condenser islocated adjacent to the turbine-generator. An overhead crane is provided for the
installation and maintenance in each turbine building. The turbine buildings are steelframed structures with insulated metal wall siding and roof decking.

Figure 8: Turbine-generator (Reproduced courtesy of NuScale Power Inc.)
Electric Power Systems
Under normal operating conditions the AC electrical power distribution system supplies
reliable and continuous power to equipment required for startup, normal operation, andshutdown of the plant. The NuScale plant does not require off-site AC electrical power
to cope with design-basis events. No backup power is required for safety system
actuation. In the event of failure of the AC electrical power supply, the DC backup
supply system provides the necessary power to ensure continuous operation of postaccident monitoring instrumentation.
9. Design and Licensing Status
The NuScale Integral System test facility is being used to evaluate design performance andimprovements, and to conduct integral system tests for NRC certification. NuScale has atarget commercial operation date of 2023 for the first plant that is expected to be built inIdaho.
10. Plant Economics
The expected cost is less than $5000 per kilowatt for such a plant, amounting to a station costof under $2.6 billion.

128

mPower (BWX Technologies, Inc.,
United States of America)
1. Introduction
The mPower™ plant consists of an integral PWR small modular reactor and related balanceof plant, designed by Generation mPower LLC to generate a nominal output of 195 MW(e)per module. In its standard plant design, 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.
Additional ‘twin pack’ plants can be added on an as-needed basis to meet demand growth.
The plant is designed to minimize emergency planning zone requirements. The mPower
reactor adopts a number of features that aim to improve the constructability, reduce overallplant complexity and enhance availability. These include: proven structure, systems andcomponents deployed commercial power plants; standard U235 fuel; modular rail and heavy
truck shippable components; fewer, smaller and simpler NSSS component designs; advanced
passive safety systems; and a two year refueling interval and conventional balance of plantsystems and components.

Figure 1: Reactor system configuration of mPower(Reproduced courtesy of Generation mPower LLC)

129

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. Development Milestones
2009
2010
2012
2014
2015
2016

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 the remaining company changed
its name to BWX Technologies, Inc., retaining its interest in 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.

4. General Design Description
Design Philosophy
The mPower design is based on the use of systems and components with an advanced plantarchitecture that reduces licensing and construction risks. The mPower design employs
passive safety features according to the defense-in-depth principle, including an undergroundsteel containment vessel structure and an underground spent fuel storage pool.
Nuclear Steam Supply System
The NSSS consists of a reactor core, a steam generator (SG), reactor coolant pumps (RCPs),
pressurizer and the core internals that are integrated within the reactor pressure vessel (RPV).
The NSSS forging diameter allows greater sourcing options and rail shipments.
Reactor Core
The reactor core consists of 69 fuel assemblies (FAs) that have less than 5% enrichment,
Gd2O3 spiked fuel rods, Ag In–Cd (AIC) control rods, and a design minimum 3% shutdown
margin. The FAs are of a conventional 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 through the use of the strong negative moderator temperature coefficientby control of the secondary side feedwater flow rates.

130

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

Generation mPower, LLC.

Country of origin

United States of America

Reactor type

Integral PWR

Electrical capacity (MW(e))

195

Thermal capacity (MW(th))

575

Expected Capacity Factor (%)> 95

Design Life (years)

60

2Plant Footprint (m )157000

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)14.8oCore inlet/exit temperatures ( C)

290.5 / 318.9

Main Reactivity Control Mechanism

Control rods

RPV Height (m)27.4

RPV Diameter (m)4.15

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine cycle

Fuel Type/Assembly Array

UO2 pellet / 17x17 square

Fuel Assembly Active Length (m)2.4

Number of Fuel Assemblies

69

Fuel Enrichment (%)<5

Fuel Burnup (GWd/ton)> 40

Fuel Cycle (months)

24

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

2

Refueling Outage (days)< 25

Distinguishing features
Modules per Plant

Internal once-through steam generator, pressurizer and control
rod drives
2

Target Construction Duration (months)

35

Seismic design

Target 85% of contiguous USA

Predicted core damage frequency (per reactor year)

Target 1E-8

Design Status

Under Development

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.

131

Reactor Coolant System
The primary cooling mechanism of the mPower reactor under normal operating condition andshutdown 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 is passively provided via the
Emergency Core Cooling System (ECCS) for continuous cooling to protect the core during asmall 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 for access to the core during refueling 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 anominal 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 undergroundwithin 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 Configuration and Approach
The inherent safety features of the reactor design include a low core linear heat rate that
reduces fuel and cladding temperatures during accidents, a large RCS volume that allows
more time for safety system responses in the event of an accident, and small penetrations athigh elevations, increasing the amount of coolant available to mitigate a small break LOCA.
The mPower plant deploys an enhanced spent fuel pool configuration, which is installed
underground, with a large heat sink to cope up for 30 days in case of loss of fuel pool cooling.
Decay heat removal system
The mPower reactor deploys a decay heat removal strategy as shown in Figure 2, with an
auxiliary steam condenser on the secondary system, water injection or cavity flooding using
the refueling water storage tank, and passive containment cooling.

132

Figure 2: mPower plant decay heat removal strategy(Reproduced courtesy of Generation mPower LLC)
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 thelarge coolant reserve provided by accumulators referred to as intermediate pressure injection
tanks (IPIT) and the in-containment refueling water storage tank (RWST), the reactor coreremains 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 waterfrom the RWST. The RWST provides a minimum of 7 days to as much as 14 days of coolingwithout the need for external intervention or AC power to maintain reactor core cooling andsafe shutdown. To address the single-failure general design criteria, the ECCS is designed forn+1 components, with all components located inside containment. There are four highpressure and four low pressure depressurization paths arranged in pairs connected to fourpressurizer connections. The ADVs are the isolation valves for the eight 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 thecore, provide long-term core cooling, and prevent the release of radioactive materials to theenvironment 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
133

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 through the 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 of steam injection during this period increases containment atmospherictemperature and pressure and quickly disperses a steam and air mixture throughout thecontainment volume. With cold walls and other structure surfaces, condensation is the mostimportant heat transfer mechanism. Natural convection and conduction-limited 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 safety systems, most notably the RWST, can extend thesetime 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 ofthe 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, thesystem is a non-safety, defense-in-depth system, providing the capacity for long-term decay
heat removal. In addition, the reactor coolant inventory and purification system (RCI) serves
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 andoperate plant systems. 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 the operators. Important operating parameters
are monitored and recorded, during both normal operations and emergency conditions toenable 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 dataprocessing capability to enable efficient operations. This layered architecture provides a highdegree of automation while ensuring safety.

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8. Plant Arrangement

Figure 3: mPower plant underground containment vessel and spent fuel storage pool
(Reproduced courtesy of Generation mPower LLC)
Reactor building
The reactor service building is a reinforced concrete, seismic Category I structure thatsurrounds the steel containment vessel and spent fuel storage pool which are located below
grade level. The control room is located below grade in the reactor service building andcontains the control system and operator interface for both reactors.
Balance of plant:
The balance of plant design (BOP) consists of a conventional power train using a steam cycle
and a water-cooled (or optional air-cooled) condenser. The BOP operation is not credited fordesign basis accidents. The steam and power conversion system is comprised of the turbinegenerator, 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 tohandle grid disconnection and is capable of 100 percent steam bypass capability to handle
turbine trip, both preventing a need for a reactor scram.
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 grid interface. Turbine-generator support systems
include a turbine bearing lubrication oil system, an electro-hydraulic control system, aturbine gland seal system, turning gear, over speed protective devices, a generator
rectifier section and a voltage regulator.
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Electric power systems
The main generator supplies power to plant auxiliaries during normal plant operation
through an isolated phase bus duct and the unit auxiliary transformer. Offsite power toplant auxiliaries during startup, shutdown, and outage conditions is supplied via backfeed from the main transformer and the unit auxiliary transformer, with the 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 todesign, license and deploy mPower modular plants. In 2013, the mPower program became
the first recipient of funding under the US Department of Energy SMR Licensing TechnicalSupport 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 specificlicensing is expected to be completed in order to support an initial deployment in the mid2020s.

136Westinghouse SMR (Westinghouse ElectricCompany, LLC., United States of America)
1. Introduction
The Westinghouse small modular reactor (SMR) is an integral pressurized water reactor(PWR) design that builds upon the concepts of simplicity and advanced passive safety
demonstrated in the AP1000® plant. The power station delivers a thermal output of 800
MW(th) and a net electrical output of greater than 225 MW(e) as a 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.

Figure 1: Reactor System Configuration of Westinghouse SMR(Reproduced courtesy of Westinghouse Electric Company LLC)
2. Target Application
The target application is the clean and safe generation of electricity; however, theWestinghouse SMR also can be used to provide process heat, district heat, and off-grid
applications, including the generation of power necessary to produce liquid transportationfuel from oil sands, oil shale, and coal-to-liquid applications.
3. Development Status
The Westinghouse SMR has been developed into a mature concept that relies heavily on
proven and licensed technology previously developed by Westinghouse during deployment ofother nuclear plants. The level of development is sufficient to perform detailed technicalevaluations and give a timely response to market demand.

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4. General Design Description
Design Philosophy
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 thehighest 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.
Nuclear Steam Supply System
The Westinghouse SMR utilizes a light water pressurized reactor coolant system (RCS) thatis integrated into a single component, which eliminates the large-break loss-of-coolant
accident (LOCA) from the postulated event.

Figure 2: Cut-Away View of the Westinghouse SMR Containment and Reactor Vessel(Reproduced courtesy of Westinghouse Electric Company LLC)
Reactor Core
The Westinghouse SMR reactor core is based on the licensed Westinghouse robust fuelassembly (RFA) design, and uses 89 standard 17×17 fuel assemblies with a 2.4 m active fuelheight and Optimized ZIRLO® cladding for corrosion resistance. A metallic radial reflector isused 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 percentof 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
surveillance intervals.
Reactivity Control
Reactivity is controlled using the Westinghouse-developed system known as MSHIM™

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control strategy or mechanical shim. MSHIM uses grey rods for short-term power control andboron dilution to adjust for fuel burnup over the longer term. Wireless instrumentation,
comprised of hardened electronics and reactor control rod drive mechanisms (CRDMs) usedin 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 provendesign 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 hightemperature-resistant, internal CRDMs for reactivity control during load-follow and similar
operations.
Reactor Pressure Vessel and Internals
The RPV and reactor internals are designed to facilitate factory fabrication and shipmentfrom the fabrication facility. Designed to meet the requirements of the American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, these components arederived from existing Westinghouse products but redesigned to function within the integral
reactor assembly. The upper internals are an integral assembly containing all of the
instrumentation and electrical penetrations to facilitate removal during refueling.
Reactor Coolant System
The Westinghouse SMR uses soluble boron in the reactor coolant for normal reactivity
depletion, and control rods for load-follow and plant shutdown. The RCS is serviced by anumber of auxiliary systems, including the chemical and volume control system, the normal
residual heat removal system, the steam generator system, the primary sampling system, theliquid and gaseous radioactive waste systems, and the component cooling water system.
Safety injection, passive boration, and heat removal are provided by the passive core coolingsystem and the ultimate heat sink (UHS) system.
Primary Cooling Mechanism
Both the integral reactor vessel and the passive core cooling system are located within thecompact, high-pressure steel containment vessel located below grade. An equipment hatch islocated in the top head of the containment vessel for maintenance activities. The containment
vessel operates at a vacuum and is designed to be fully submerged in water to facilitate heatremoval during accident events while providing an additional radionuclide filter.
Steam Generator / Pressurizer / Reactor Coolant Pumps
The Westinghouse SMR design incorporates eight seal-less canned motor pumps, which aremounted horizontally to the shell of the RPV just below the closure flange to provide forced
reactor coolant flow through the core. A central primary riser directs the coolant flow as itexits the core to the steam generator. The reactor vessel downcomer acts as the channel fordelivering the coolant flow from the reactor coolant pumps to the core inlet. The steam
generator utilizes straight tubes with the primary reactor coolant passing through the inside of
the tubes and the secondary coolant passing on the outside. An integral pressurizer is locatedabove 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 located outside of containment, reducing the reactor andcontainment vessel heights by approximately 6 meters. The steam generator/pressurizer
assembly can be removed for refueling operations through a bolted closure flange near the
top of the integral reactor vessel.

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MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Westinghouse Electric Company LLC

Country of origin

United States

Reactor type

Integral PWR

Electrical capacity (MW(e))˃ 225

Thermal capacity (MW(th))

800

Expected capacity factor

95

Design life (years)

60

Plant footprint (m2)65000

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure

15.5oCore inlet/exit temperatures ( C)294/324

Main reactivity control mechanism

CRDM, boron

RPV height (m)28

RPV diameter (m)3.7

Configuration of reactor coolant system

Integral type

Power conversion process

Indirect Rankine Cycle

Fuel type/assembly array

UO2 pellet/17×17 square

Fuel assembly active length (m)2.4

Number of fuel assemblies

89

Fuel enrichment (%)<5

Fuel burnup (GWd/ton)˃ 62

Fuel cycle (months)

24

Cogeneration capability

Possible

Approach to engineered safety systems

Passive

Number of safety trains

Three diverse decay heat removal

Refuelling outage (days)

17

Distinguishing features

Incorporates passive safety systems and proven componentsof the AP1000 plant and earlier Westinghouse designs

Modules per plant

1

Target construction duration (months)

18–24

Seismic design

Based on CEUS sites

Core damage frequency (per reactor-year)< 5E-8Design Status

Concept design completed

5. Safety Features
The Westinghouse SMR is an advanced passive plant where the safety systems are designed
to mitigate accidents through the use of natural driving forces such as gravity flow andnatural circulation flow. The plant is not reliant on alternating current (ac) power or othersupport 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 applied in theoperating plants. The integral reactor design eliminates large loop piping and potential large140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 of an accident relative to alarge 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 is a plant capable of withstanding natural phenomena hazards and beyond-design-basis
accident scenarios, including long-term station blackout.
Engineered Safety System Configuration and Approach
The Westinghouse SMR is designed with passive safety systems that utilize the natural forcesof evaporation, condensation, and gravity. The design basis and licensing of passive systems
were first implemented in the design of the AP1000 plant. Elements of these systems aredescribed in the following sections.
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 minutes of 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 UHS
system tanks. The UHS tanks are sized to provide a minimum of 7 days of decay heatremoval, 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 withdiverse bleed-and-feed methods, including a two-stage automatic depressurization systemthat 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 and condensed by the containment shell. The containment shell is cooled by thewater 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 thetwo redundant UHS tanks that maintain the OCP full of water. The water condensed on thecontainment shell flows back into the RCS through two sump injection flow paths.

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Figure 3: Three Diverse Decay Heat Removal Methods of the Westinghouse SMR(Reproduced courtesy of Westinghouse Electric Company LLC)
Containment System
The containment vessel is a carbon steel vessel that is normally submerged in a pool of water.
Pressure in the containment vessel following postulated events is maintained by transferring
heat through the shell to the water surrounding it. As this water boils, the inventory is madeup from the two large UHS tanks that supply the plant with enough decay heat removal
capacity for more than 7 days.

Figure 4: Below-Grade Location of Westinghouse SMR Reactor and Containment Vessels(Reproduced courtesy of Westinghouse Electric Company LLC)
6. Plant Safety and Operational Performances
The design of the Westinghouse SMR represents a significant advancement in plant safetywith an estimated core damage frequency of 5E-8 per reactor year while maintaining an
expected capacity factor of 95 percent.

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7. Instrumentation and Control Systems
An Ovation™-based digital instrumentation and control (I&C) system controls the normal
operations of the plant. The protection and safety monitoring system provides detection of
off-normal conditions and actuation of appropriate safety-related functions necessary toachieve and maintain the plant in a safe shutdown condition. The plant control system
controls non-safety-related components in the plant that are operated from the 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 trademarks of theirrespective owners.

8. Plant Arrangement
Reactor Building
The entire nuclear island is divided into five distinct and physically separate sectors all
located below grade. The five independent sectors house a radiological control area that
includes the reactor vessel, high-pressure containment vessel and spent fuel pool, and allsafety-related systems, including safety-related dc power and I&C systems. No piping,wireways, or passageways traverse between these independent sectors. Access to each of thefive independent sectors is instead regulated above grade. This design provides protection
against external threats and natural phenomena hazards.

Figure 5: Plant Layout of Westinghouse SMR(Reproduced courtesy of Westinghouse Electric Company LLC)
Reactor Control Room
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:
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 toaccommodate a wide variety of backpressures with different blade configurations
optimized for narrow-range, high-performance power. The water intake requirements
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will be comparable to existing plants on a per-power basis, but significantly less on aplant 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 adc 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 Class1E and non-Class 1E dc power systems. Safety-related dc power isprovided to support reactor trip and engineered safeguards actuation. Batteries are sized
to provide the necessary dc power and uninterruptible ac power for items such asprotection 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) dieseldriven generators to provide power for defence-in-depth electrical loads, and 2) a decay
heat-driven generator. The decay heat-driven generator provides ac power to the plantusing the heat 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. andUK licensing. Westinghouse is considering a number of business models for the successful
deployment of the Westinghouse SMR product globally. In addition, in February 2015, theU.S. Nuclear Regulatory Commission (NRC) approved Westinghouse’s testing approach for
the Westinghouse SMR design. The NRC approval is a significant step toward design
certification and will reduce the time ultimately needed to license the Westinghouse 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 frominside and outside of Westinghouse, identified what would occur in the unlikely event of asmall-break LOCA in the Westinghouse SMR. It also defined the test program thatWestinghouse will conduct in the future to prove that its safety systems would safely shut
down the reactor in response to a small-break LOCA. As a major technical innovation, thepotential 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. Plant Economics
Given that some elements of nuclear power plant design come with a relatively fixed cost
regardless of power output, economy of scale challenges all SMR designs. To address thechallenge, the Westinghouse design maximizes the power output for the material used in thedesign. Additionally, it is recognized that a significant contributor to the cost of nuclear isassociated with time required to construct the plant and the cumulative cost of money duringthat time. To condense the construction schedule, the Westinghouse SMR is fully modular.
Structures, systems, and components are manufactured in fabrication facilities and shipped tothe site for assembly, allowing for extensive parallelization and schedule compression.

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SMR-160 (Holtec International,
United States of America)
1.

Introduction

The SMR-160 conceptual design has been developed by Holtec International as an advanced
PWR-type small modular reactor producing power of 525 MW(th) or 160 MW(e) adopting
passive safety features. Simplification in the design is achieved by using fewer valves,
pumps, heat exchangers, instrumentation, and control loops than conventional plants,
simplifying operator actions during all plant modes, including diagnosing and managing offnormal and accident conditions. The SMR-160 uses fuel similar to existing commercial LWR
fuel, includes no reactor coolant pumps and utilizes a large vertical steam generator. A
modular construction plan for SMR-160 involves pre-assembling the largest shippablecomponents prior to arrival at a site. A 24-month construction period is envisaged for each
unit.

Figure 1: SMR-160 Nuclear Power Plant (Reproduced courtesy of Holtec International, Inc.)
2. Target Application
The primary application of SMR-160 is electricity production with optional cogenerationequipment (i.e., hydrogen generation, district heating, and seawater desalination). Target
applications include distributed electricity production, repowering coal facilities, uprate forexisting nuclear facilities and providing electricity and low temperature process heat forcommercial and military installations. Design optimization includes air cooled condensation
for no wet cooling. The plant design includes a unique fuel management solution thatincorporates an integrated state-of-the-art underground dry fuel storage facility with the baseplant offering.

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3. Development Milestones
April 2015

Conceptual Design Complete

November 2017

Preliminary Design Complete

April 2018

PSAR/PSID Complete

4. General Design Description
Design Philosophy
The SMR-160 design is driven by the principal criterion that all safety significant systems
must be powered by natural circulation. Core strengths and innovation of SMR-160 are its
inherent safety, security, constructability and simplicity of operation.

Figure 2: Detailed view inside SMR-160 containment(Reproduced courtesy of Holtec International, Inc.)
Nuclear Steam Supply System
The SMR-160 is a pressurized water reactor with reactor coolant system (RCS) using natural
circulation for all power and accident modes and states. The RCS is comprised of the reactor
pressure vessel (RPV) and a steam generator (SG) in an offset configuration with an
integrated pressurizer flanged to the top of the steam generator. The RPV and the SG areconnected by a single connection which houses both the hot leg and the cold leg. The hot leg
is the inner pipe and the cold leg is the annular region of this single connection. Uniqueamong integrated PWRs, the offset configuration allows easy access to the core without
moving the RPV or SG during refuelling. The SG has a superheating feature which eliminates
the need for a moisture separator reheater and trains of feedwater heaters, while notcompromising the thermodynamic efficiency of the plant. The secondary side has only two
feedwater heaters simplifying plant operations and maintenance146

Reactor CoreThe SMR-160 employs an efficient core design that uses traditional reload shuffle at the endof each cycle. The core is designed for up to a two year cycle with flexibility for shorter
cycles depending upon utility requirements. Because of the relatively low number of fuelassemblies and overall outage schedule, the shuffling of nuclear fuel is not a critical pathactivity during an outage. The core contains an array of 96 square-lattice assemblies that areapproximately 5.5 inches square and 14 ft. long in active height. Fuel bundles are surroundedby 16 reflector assemblies that contain pins of solid stainless steel or other reflecting material
contained within stainless steel tubes. The reflector assemblies allow for an axial variation in
the radial reflector by varying the composition of the pins along their length. The reflector
assemblies can also be changed on a cycle to cycle basis if necessary depending upon fuelutilization requirements. Surrounding all of the fuel and reflector assemblies is a solid
stainless steel reflector structure. This provides reflection of fast neutrons in order toimprove the neutron economy, while reducing the fast neutron fluence through the reactor
pressure vessel thereby, increasing vessel life.
Reactivity Control
Burnable absorbers integral to the fuel are used for long-term reactivity control and aredesigned to optimize 3D power distributions, cold shutdown margin, and hot excess
reactivity. Short term changes in reactivity are controlled by movements of cruciform control
blades, eliminating the need for using boric acid as a chemical shim. A core operated withoutchemical shim reduces operational costs and minimizes radioactive waste. In addition,boron-free operation increases reactor stability due to a strong negative moderator
temperature coefficient of reactivity. The 21 cruciform control blades are very similar to
BWR blades in that they contain an array of stainless steel tubes in each wing. The stainlesssteel tubes are filled with compacted B4C powder that serves as the thermal neutron absorber.
The control blades enter from the top of the RPV, eliminating penetrations in the bottom
vessel head.
Reactor Pressure Vessel and Internals
The RPV is an ASME code III, Class 1, thick-walled cylindrical pressure vessel with an
integrally welded bottom head and a removable top head. The upper extremity of the RPV
shell is equipped with a tapered hub flange, which is bolted to a similar flange welded to the
top head. The offset configuration of the SG and RPV also enables the use of traditional
external control rod drive mechanism.
Reactor Coolant System
The SMR-160 RCS operates purely by gravity, as reactor coolant circulates entirely by virtueof the density difference in the primary water, which is present as long as fuel assemblies in
the core produce heat. There are no reactor coolant pumps in the system. The RCS consists ofthree major components as shown in Figure 4: the (1) RPV and (2) SG with integral (3)Pressurizer.
Steam Generator
The SMR-160 has a vertical once through straight tube SG with the reactor coolant flowing
inside Inconel 690 thermally treated tubes. The vertical straight tubes provide easy access forin-service inspection of the tubes. The SG uses sub-cooled feedwater to produce superheatedsteam on the shellside. The SG features a large inventory of secondary water on the shellsidewhich provides substantial time margin to dryout (>4 hours).

147

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Holtec International

Country of origin

USA

Reactor type

PWR

Electrical capacity (MW(e))

160

Thermal capacity (MW(th))

525

Expected capacity factor

˃ 98

Design life (years)

80

Plant footprint (m2)20500

Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure

15.5oCore inlet/exit temperatures ( C)

196/316

Reactivity control mechanism

CRDMs

RPV height (m)15

RPV diameter (m)2.7

RPV or module weight (metric ton)

200

Configuration of reactor coolant system

Integral type

Power conversion process

Indirect Rankine Cycle

Fuel type/assembly array

UO2 pellet/ square array

Fuel assembly active length (m)3.7

Number of fuel assemblies

N/A

Fuel enrichment (%)<4.95

Fuel burnup (GWd/ton)

45 GWd/t (initial design)

Fuel cycle (months)

24 months

Cogeneration capability

Possible

Approach to engineered safety systems

Passive

Number of safety trains

2

Refuelling outage (days)

10

Distinguishing features•

Modules per plant••
1

Unique approach to Defence-in-Depth with active nonsafety and passive safety cooling systems
Integrated, indefinite, passive spent fuel coolingIntegrated spent fuel dry storage

Pressurizer
The pressurizer is integral to the steam generator and engineered to maintain primary waterand vapor interface to control pressure in the RCS. The pressurizer contains conventional
heating and quenching elements (water/steam).
5. Safety Features
The SMR-160 safety basis incorporates defence-in-depth via multiple and diverse simple
pathways for heat rejection from the core. All safety systems are within the containment
structure, making them secure and safe from external threats. A large inventory of water148

within a reservoir outside the containment structure provides long-term post-accident coping,
and that structure facilitates air cooling for decay heat removal if the reservoir inventory boils
off; for an unlimited coping period. SMR-160 engineered safety features include the passive
core cooling system (PCCS), passive spent fuel pool cooling system (PSFPCS) and automatic
depressurization system (ADS). A Holtec International innovative design; the PCCS andPSFCS are tied to the integrated containment system (ICS), which is the plant ultimate heatsink (UHS).

Figure 3: SMR-160 Integrated Containment System(Reproduced courtesy of Holtec International, Inc.)
Integrated Containment System
The SMR-160 ICS consists of a free standing steel containment vessel called the containment
structure (CS), supported within a reinforced concrete reactor building called the containment
enclosure structure (CES), which also provides missile protection (See Figure 3). An annular
coolant reservoir (CR) between the CS and the CES is filled with water and serves as the
UHS for SMR-160. In addition to preventing the release of radioactive fission products to theenvironment, the ICS acts as a large passive heat exchanger to cool the core and the spent
fuel pool. The PCCS and PSFPCS use this system (via the CS metal surface) to reject heat tothe environment. The reservoir provides sufficient inventory to remove the peak decay heat;as the water in this reservoir depletes (reservoir becomes dry > 60days), the ICS transitions to
air cooling to remove subsequent decay heat. The operator has the option of replenishing thereservoir. The SMR-160 defence-in-depth architecture incorporates non-safety active systems
for core and spent fuel pool cooling. The safety systems and non-safety systems are also
configured to mitigate severe accidents.
Engineered Safety System Configuration and Approach
In the event of abnormal transients or postulated accidents, SMR-160 will employ non-safety
active cooling systems as a first line of defence. Failure of those systems will actuate thepassive safety systems and the plant will shut down and remain safely cooled for an unlimited
period without the need for power, make-up water or operator actions, eliminating the needfor safety related emergency diesel generators (EDG). Engineered Safety Systems are shownin Figure 4. The PCCS consists primarily of two sub-systems - the submerged bundle coolingsystem (SBCS) and the passive core makeup water system (PCMWS).

149

Figure 4: Schematic representation of SMR-160 engineered safety systems(Reproduced courtesy of Holtec International, Inc.)
Submerged Bundle Cooling System
For non-Loss-of-Coolant-Accident events (non-LOCA), the SBCS is actuated. This system isprincipally comprised of two heat exchangers – namely the submerged bundle heat exchanger(SBHX) and the heat dissipation Ducts (HDDs). In the event of an accident, primary coolant
is routed to the SBHX, which is functionally equivalent to an auxiliary steam generator. The
steam produced by this heat exchanger is routed to the CS via the HDDs, where it condenses
by rejecting the latent heat to the containment reservoir through the steel CS. This system
uses the elevation difference between the reactor and the SBHX to drive natural circulation
flow of the primary coolant, and the elevation difference between the SBHX and HDDs todrive natural circulation flow of the secondary coolant. The SMR-160 also features an active
non-safety residual heat removal system (RHR) for normal decay heat removal.
Passive Core Makeup Water System
In the event of a Loss-of-Coolant-Accident (LOCA), the PCMWS provides coolant inventory
to keep the core covered at all times. The system provides medium pressure coolant injection
using accumulator tanks (pressurized with nitrogen) for short term inventory make up andgravity injection for long term cooling. The coolant inventory lost via the break flashes into
steam and liquid. The steam rises up and mixes with the containment atmosphere andcondenses on the inside wall of the CS by rejecting its latent heat to the CR. The condensate
and the entrained liquid are collected in the recirculation sump via gutters. Once the RCS
reaches containment pressure, the coolant inventory in the sump reaches the core via gravity(gravity injection) due to the elevation difference between the sump and the core. For small
break LOCA, the ADS actuates and the depressurization valves open (introduces anadditional break) in order to depressurize the RCS, allowing gravity injection to keep the core
covered. Gravity injection will commence once the RCS pressure equalizes with thecontainment pressure. In both sub-systems, the ICS is utilized as the heat exchanger(Ultimate Heat Sink) to passively reject decay heat to the environment.
Passive Spent Fuel Pool Cooling System (PSFPCS)The SMR-160 spent fuel pool is located within containment. This feature not only contains
150radioactive release from the spent fuel but also passively cools the spent fuel by tying to the
ICS. By incorporating interim dry storage within the plant design, the spent fuel inventory in
the plant never exceeds 250 assemblies. The plant design features both passive and active
non-safety pool cooling systems.
6. Plant Safety and Operational Performance
The SMR-160 natural convection-driven reactor coolant loop is coupled with an optimized
simple steam cycle, and is well adapted to load following
7. Instrumentation and Control systems
SMR-160 utilizes the Mitsubishi Electric Total Advanced Controller (MELTAC) platform as
the backbone for the plant’s I&C/HSI design. MELTAC – now proven with over 300 years ofoperating reactor experience - provides exceptional performance with attributes that facilitatea highly integrated plant-wide digital control system (DCS) configuration. MELTAC provides
unique nuclear specific I/O and configuration flexibility to perform all nuclear safety andnon-safety functions using the same digital platform. This includes Reactor Trip andEngineered Safety Feature actuation; integrated control functions such as feedwater, turbine
and rod control; and I/O modules directly compatible with in-core and ex-core neutron
monitoring detectors, and radiation monitoring detectors. The SMR-160 I&C/HSI minimizes
plant capital cost, as well as long term operation and maintenance costs, by minimizing thenumber of platforms (two), safety divisions (two), controllers, sensors, processors, field
cables, operators (two operators with the capability to operate multiple units from a single
control room), and plant staff by eliminating most periodic safety system surveillance andincluding on-line condition monitoring. The HSI features a Large Display Panel (LDP) - a
spatially dedicated continuously visible (SDCV) overview display of critical safety andpower production functions. The HSI also includes a Safety Console (SC); Class 1E LCD
touch screens, referred to as safety visual display units (S-VDU), provide SDCV display TypeA (support for credited manual actions, if any) and Type B (critical safety functions)parameters. The operator console (OC) is comprised of safety grade and non-safety grade
LCD touch screens; non-safety operational video display units (O-VDU) are multi-divisiondisplays with soft controls for all plant functions and components (non-safety and safety). SVDUs are used to control all safety functions and components. I&C architecture includes a
Plant Safety System (PSS), Plant Control System (PCS), Diverse Actuation System (DAS)and Information Technology System Interface (ITS). PSS elements include the Reactor
Protection processor, Component Control processor, Safety Bus and S-VDU processor. A
plant IT System interface for operators is accommodated in the control console design. Data
from the PSS, DAS, and PCS is sent to the plant’s IT System network via the PCS Gateway
to support long term archive, O&M applications, and corporate wide information
accessibility.
8. Plant Arrangement
Reactor building
The SMR-160 reactor building shares a common basement with the auxiliary building.
Nearly half of the reactor building is underground and the reactor core is located at the verybottom of the ICS, protecting it from unauthorized access and missile impact. Figure 5
provides a depiction of the plant layout.

151

Figure 5: Plant Layout of SMR-160 (Reproduced courtesy of Holtec International, Inc.)
Control building
The SMART compact control room is designed for two man operation (staffing goal) undernormal conditions of the plant. The main control room (MCR) is designed to provide air andcooling passively in any emergency situations.
Balance of plant:
Steam is produced at a relatively low pressure, 335 psi, but with considerable superheat,
170°F. The low pressure steam passes through an intermediate-low pressure turbine. The
plant features only two feedwater heaters to supply sub-cooled feedwater to the steam
generator.
Turbine Generator building
SMR-160 features an axial/side exhaust steam turbine with an option to use air cooled
condensation. Superheated steam produced by plant is conducive for use in non-electric
applications and also co-generation plants.
Electric power systems
The SMR-160 electric power system consists of the main generator, main transformer
auxiliary transformers, non-safety diesel generators and safety related 1E batteries.
There are two trains of Class 1E and non-Class 1E power supply. SMR-160 Class 1E
power is required only for one-time alignment of valves for safety features actuationand post-accident monitoring. The Class 1E power supply derives power fromphysically separate battery banks.
9. Design and Licensing Status
Pre-application engagements are taking place with domestic and international regulators, and
the Preliminary Safety Analysis Report is scheduled for completion by early 2018.
10. Plant Economics
Confidential plant economic data (capital cost, LCOE) have been developed and shared with,and are available to, interested commercial clients and stakeholders.

152WATER COOLEDSMALL MODULAR REACTORS(MARINE BASED)154

ACPR50S (CGN, China)
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 potential optimal solution for
combined supply of heat, electricity and fresh water for marine resource development
activities, energy supply and emergency support on islands and along the coastal area.

Figure 1: Reactor System Configuration of ACPR50S(Reproduced courtesy of CGNPC)
2. Target Application
Being an offshore floating SMR, the ACPR50S is designed as a multipurpose power reactor
for the following applications: combined energy supply for offshore oil drilling platform;
offshore combined energy supply; coastland and island combined energy supply; energy
supply for offshore mining, nuclear power ship; and distributed clean energy portfolio
together with solar energy and wind power for islands. The ACPR50S has two (2) energy
supply modes i.e. heat-electricity-fresh water-steam cascade supply mode and combined heatelectricity-fresh water-steam supply mode.

Figure 2: ACPR50S Energy Cascade Supply(Reproduced courtesy of CGNPC)

155

Figure 3: ACPR50S for floating and fixed platform(Reproduced courtesy of CGNPC)

Figure 4: ACPR50S for offshore combined energy supply station(Reproduced courtesy of CGNPC)
3. Development Milestones
2009 – 2011

Conceptual study

2012

Starting Conceptual Design and formulating plan of theoretical tests

2013

Completion of Conceptual Design to determine the plan and major parameters

2014

Completion of Overall Design, preliminary demonstration and theoretical tests of the key
systems; R&D on software; Starting study of Codes and Standards;
Preliminary Design Stages

2015 – 2016
2015
2016
2017 – 2020
2017
2018
2020 ~

156

Completion of imported tests; completion of design software suitability evaluation; starting
Preliminary Design; Completion of Codes and Standards study
Completion of Preliminary Design and Preparation of Construction; Starting key equipment
procurement for Project Construction Stage
Project Construction Stage
Completion of PSAR review; Obtaining construction permit; Starting construction; and starting
construction design
Submission of FSAR and EIA; Completion of construction and installation; and obtaining fuelloading permit
Start up commissioning and connection to grid

4. General Design Description
Design Philosophy
The ACPR50S adopts design simplification with less cost and lower investment risks in order
to be competitive with conventional offshore energy sources. Modular design is adopted
through standardized streamline manufacturing aiming for shorter construction period as wellas less cost. Higher load factor to be attained by a long refuelling cycle. ACPR 50S iscapable of providing safe, clean, stable and efficient energy to satisfy various demands onheat, electricity, water and steam.
Nuclear Steam Supply System
The compact loop-type PWR nuclear steam supply system (NSSS) design of the ACPR50S
consists of the reactor pressure vessel(RPV) that houses the core, two once-through steam
generators (OTSG), two main coolant pumps(RCP) and a pressurizer(PZR) that areinterconnected by short reactor coolant system legs. The compact layout of first loop reduces
sharply the probability of LOCA. The primary cooling system is based on forced circulation
during normal operation. The system has natural circulation capability under certain
conditions, for example, under the condition of 10% thermal power and 14% flow rate, thenatural circulation can be established.
Reactor Core
The low power density design with a slightly enriched UO2 fuelled core ensures a thermal
margin of greater than 15% which can accommodate any anticipated transient event. Thisfeature ensures the core thermal reliability under normal and accident conditions. The fuelassemblies (FAs) of ACPR50S core with an axial length of 2.2 m core have a squared 17x17
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.
Reactivity Control
The reactivity is controlled by means of control rods, solid burnable poison and soluble boron
dispersed in the primary coolant. Burnable poison rods are introduced for flat radial and axial
power profile, which results in an increased thermal margin of the core. The number andconcentration of the burnable absorber rods in each fuel type are selected so that reactivity ofeach assembly can be as flat as possible. There are 16 control rods, with a magnetic force
type control rod driving mechanism (CRDM).
Reactor Pressure Vessel and Internals
ACPR50S reactor pressure vessel envelopes and holds the core and the RPV internals, so thatthe fission reaction of the nuclear fuel is limited in one space. The RPV internals are used tosupport and fix the core component.
Reactor Coolant System
The ACPR50S primary cooling mechanism under normal operating condition and shutdown
condition is done by forced circulation. The RCS has been designed to ensure adequatecooling 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,
so 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
area of postulated small break LOCA.

157

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

China General Nuclear Power Group (CGNPC)Country of origin

China

Reactor type

Loop type PWR

Electrical capacity (MW(e))

60Thermal capacity (MW(th))

200

Expected Capacity Factor (%)

95

Design Life (years)

40

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)

15.50
o

Core inlet/exit temperatures ( C)

299.3/ 321.8

Main Reactivity Control Mechanism
RPV Height (m)Control rod driving mechanism (CRDM), solid burnable poison andboron solution
7.2

RPV Diameter (m)2.2Module Weight (metric ton)

380

Configuration of Reactor Coolant System

Compact Loop (with short NSSS connecting pipes)

Power Conversion Process

Indirect Rankine cycle

Fuel Type/Assembly Array

UO2/17x17 square pitch arrangement

Fuel Assembly Active Length (m)2.2

Number of Fuel Assemblies

37

Fuel Enrichment (%)<5%

Fuel Burnup (GWd/ton)< 52

Fuel Cycle (months)

30

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Passive

Number of safety trains

2

Refueling Outage (days)

7 (to be defined later)

Distinguishing features
Modules per Plant

Floating power boat, once-through steam generator, passive safetysystemSingleTarget Construction Duration (months)

36

Predicted core damage frequency per reactor year

< 1E-6Large release frequency per reactor year

< 1E-8Design Status

Completion of conceptual/program Design, Preparation of project
design158

Steam Generator
The ACPR50S has two OTSGs with helically coiled tubes to produce superheated steam
under normal operating conditions. The OTSGs are located on both sides of the reactor
vessel. The small inventory of the secondary side (tube side) water in each OTSG prohibits areturn to power following a steam line break accident. In the case of accidents, the OTSG canbe used as the heat exchanger for passive residual heat removal system (PRHRS), which
permits an independent operation of the PRHRS regardless of the hydraulic condition of the
primary system.
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 thesystem pressure at nearly constant level for the overall plant operation. The primary systempressure is maintained nearly constant due to the large pressurizer steam volume and theheater control. As the volume of the pressurizer is designed sufficiently large, condensing
spray is not required for the load manoeuvring operation. The reactor over-pressure at thepostulated design basis accidents related with a control failure can be reduced through theactuation of the pressurizer safety valve (PSV).

Figure 5: ACPR50S Reactor Building inside the Ship
(Reproduced courtesy of CGNPC)
5. Safety Features
The ACPR50S is designed with enhanced safety standards from the Generation III reactor
designs to meet the requirements of national laws and regulation on the environmental releaseor radioactivity. A compact NSSS configuration with short nozzles will lead to reduced
probability of LOCA. The NSSS in the ship is located under sea water level. The seawater isused as the ultimate heat sink and radiation shielding. Severe accident mitigation measures
are incorporated to ensure low radioactive substance release probability to eliminate need foroff-site emergency response. The ACPR50S is designed to cope with extreme external events
such as typhoon, tsunami and ship collision.

159

Engineered Safety System Approach and Configuration
The ACPR50S is designed with passive safety systems which 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 exhaust system (CHE).Thesepassive safety systems are used to cope with design basis accidents (DBA) and severe
accidents with core melts. The engineered safety system of the ACPR50S is shown in
Figure 6.

Figure 6: Engineered safety system of ACPR50S (Reproduced courtesy of CGNPC)
Decay Heat Removal System
The SHR is developed to remove the decay heat if the normal decay heat removal pathway isunavailable 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. And the
decay heat is removed to the cooling water in the cooling water tank outside the containment
eventually. In case of power supply is still available, if the normal decay heat removal
pathway fails under non LOCA accident, the feed water system will be started instead of the
SHR.
Emergency Core Cooling System
The SIS is a very important engineered safety system which is developed to cool the core andboride following DBA and extends the no operating time of plant operators to 7 days. Its
main function is to control and mitigate the consequences of the accident to prevent the DBA
from becoming serious accident t beyond design basis accident (BDBA). The core cooling iscompletely driven by the natural force following the DBA such as LOCA, which makes thesystem's composition and operation greatly simplified. Both high pressure and low 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, middle pressure and low
pressure safety injection, ADS is developed.
160

Containment System
The containment system is to contain radioactive material and to protect the environment
against primary coolant leakage. The containment system of ACPR50S is actually a cabin
called reactor cabin, a square containment. The containment has a volume of 870 cubic
meters, a design internal pressure of 1.4 MPa and a design external pressure of 0.3MPa.The
CHE is used to reduce the concentration of hydrogen in the containment to the safety limit
under DBA and BDBA and to continuously monitor the hydrogen concentration at the top of
the containment. The CPS is developed to cope with the DBA which can lead to the pressure
rise inside containment vessel and suppress the containment pressure peak to ensure theintegrity of the containment. The CPS is composed of the suppression pool system andsuppression pool cleaning and cooling system. The containment heat removal system (CHR)
is used to prevent the containment slow over temperature and overpressure. The CHR is apassive natural circulation system, and can provided 7 days of cooling for the containment in
the case of no external water apply. The containment isolation system (CIS) is to provide
isolation for the containment and to prevent and restrict the escape of radioactive fission
products in the event of an accident.
6. Plant safety and Operational Performances
ACPR50S is suitable for all kinds of sea states such as swing, concussion, vibrancy, typhoon,
seaquake and so on. The operational power range can be justified from 50% to 100% Pn(Power (nuclear)) and can be operated steadily for a long period for any power level of 50%
to 100% Pn in order to satisfy the needs of users.
7. Instrumentation and Control systems
The I&C system design for ACPR50S will be based on defence in depth concept, compliancewith the single failure criterion and diversity. The diversity in the design of I&C system isachieved through: The I&C system design for ACPR50S is mainly used in the steady-state
and transient power during operation and provides automatic protection against unsafe
reactors and abnormal operation, and provide the trigger signal to mitigate the consequencesof accident conditions. Two reactors share one control room and one technical support center,a separate two remote shutdown stations, to ensure control and operation of the plant undernormal and accident conditions.
8. Plant Arrangement
The single reactor module with the electrical, the steam generator, and nuclear auxiliary
facilities is installed inside a non-propelled barge mounted ship for sharing facilities andreduced cost. Plant main building consists of the reactor containment cabin, the nuclear
auxiliary cabin, the emergency diesel generator cabin and the turbine-generator cabin. For
efficient radiation management, the plant main building is sub-divided into two zones, theduty zone and the clean zone. Systems linked with refuelling, overhauling, radwaste
treatment are installed in the onshore basement.
Floating platform
The reactor building and fuel storage area are equipped with a full monitoring system withclosed circuit monitoring system to oversee and prevent unauthorized access to the fuel.
Reactor building is a pre-stressed concrete shell structure composed of a right cylinder with ahemispherical 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 and fuel storage area are located. Figure 7 shows the overview of the reactor
cabin, auxiliary cabin, turbine-generator cabin, the emergency diesel generator cabin and the161main control room arrangement of ACPR50S.
Onshore basement
The onshore basement of ACPR50S houses the fuelling building, the radwaste treatmentbuilding, and other balance buildings of plant. Refuelling and overhauling will be performed
in onshore basement.

Figure 7: ACPR50S general plant arrangement in the float platform(Reproduced courtesy of CGNPC)
Control room
The compact control room is designed for one man operation under normal conditions of theplant. 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 and principles are applied and verified using the full scope dynamic mock up.
9. Design and Licensing Status
The ACPR50S has completed the conceptual/program design work and is now preparing for
project design. The National Nuclear Safety Administration（NNSA） is enacting the rule ofsafety review for small modular reactor. The ACPR50S plans to apply for the construction
permit to the NNSA in 2017. An industrial demonstration plant of ACPR50S is being planned
to be constructed in China, with a target of startup commissioning in 2020.
10. Plant Economics
The target plant construction cost of ACPR50S is about $7900/kWe and an electricity
generating cost of ¢19/kWh.

162

Flexblue (DCNS, France)
1.

Introduction

Flexblue is a transportable and immersed nuclear power system (TINPS), with an output
capacity of 160 MW(e). The system is designed to deliver electricity using a Transportable &Immersed Nuclear Unit (TINU), later called “module” that is anchored with a positive
floatability on the sea bed, between 40 m and a maximum depth of 100 m. Each module has
the length of 146-meters and 14-meter diameter. Additional modules can be installed as
demand increases. The nuclear units are positioned close to the shore and must be within the
territorial waters. The reactor section is sealed in this configuration. A module encompasses areactor (PWR with fuel enrichment classically less than 5%) and a turbine generating theelectricity: it is manufactured in an assembly-line and assembled in a shipyard. System
control and connection to the power grid are grouped together in a land-based facility called
the “Shore Control Center” (SCC). They are connected to the immersed modules (or the farmof modules) via underwater buried cables. Fuel unloading and reloading and major
maintenance operations are performed on a coastal base called the “Service Factory”(Supporting Site), which may be located at significant distances from the Operating Site (inthe same country or not). Figure 1 shows reactor system configuration.

Figure 1: Reactor system configuration of Flexblue (Reproduced courtesy of DCNS)
2. Target Applications
Flexblue is designed to supply electricity to coastal grids. Figure 2 shows the life cycle andarrangement of Flexblue that comprises two sites for Operating Site and Service Factory. The
production cycle duration is equal to 3 years (in the Operating Site) and the theoreticalavailability over the 60 years of life-time is equal to 91%. After each cycle, the module istransported back to the Service Factory for major maintenance and refueling. Every ten years,
a detailed inspection is also performed on this site, according to the international
requirements. The Service Factory can be used and mutualized for many modules, whatever
their owner and Operating Site origin.

163

Country of
OriginService Factory
Assembly Line& Shipyard

Without Fuel Module
Transportation

Fuelled Module
Transportation

Operating Site

Figure 2: Flexblue Life Cycle and Arrangement (Reproduced courtesy of DCNS)
The following scenario is retained for the purpose of this document: these two installations
are each located on the territory of two distinct countries. The life-cycle steps follow asshown in Figure 2; steps 4 – 5 – 6 – 7 – 8 are performed for the 60 years of the nuclear unit
life-time:
1 – The module is assembled within the shipyard. Preliminary testing is performed withoutfuel.
2 – The module is transported to the Service Factory without any fuel.
3 – First fuelling and module trials are performed in the Service Factory. The first criticality
is committed at that time.
4 – Module transportation with fresh fuel is carried out toward the Operating Site, throughout
the international sea.
5 – Nautical operations are performed to position and to anchor the module on the sea-bed.
6 – The module delivers power to the grid via the Shore Control Centre. Minor maintenanceis performed for the production operations in the Operating Site.
7 – Module transportation with spent fuel is carried out toward the Service Factory,
throughout the international sea.
8 – It corresponds to the module’s outage. Irradiated fuel is put within a devoted spent fuelpool in the Service Factory site and refuelling is carried out. Waste is removed from themodule and major maintenance operations are performed.
By design (except when testing), the modules reactivity is authorized only when it isanchored and connected to the grid. Under all other circumstances, the reactor is shut down
and criticality made impossible.

1643. Development Milestones
2012 – 2016

The Project gathers international financial and operator stakeholders.

4. General Design Description
Design Philosophy
Flexblue design adopts proven technology that draw on DCNS’s experience in nuclear
propulsion and submarine power plants. The reactor design coupled with the concept ofmooring it 100 m below the sea surface is envisaged to ease some of the safety issues of landbased nuclear power plants. The reactor is perceived to be resistant to natural disasters suchas earthquakes, tsunamis and floods. The reactor will have the advantages of the ocean as a
built-in coolant and ultimate heat sink.
Nuclear Steam Supply System
The module encompasses a 2-loop PWR with two horizontal steam generators. The Nuclear
Steam Supply System (NSSS) operation diagram is presented below, in relation with thediverse configurations of the module.

Figure 3: Reactor Operation Diagram of the module’s NSSS
(Reproduced courtesy of DCNS)
Reactor Core
The reference core is made of 77 classical 17x17 fuel assemblies with an active length of
2.15 m. Enrichment is kept below 5% and reactivity is controlled without soluble boron.
This latter characteristic reduces the generation of radioactive wastes and simplifies thechemical control system. Flexblue power production cycle lasts 38 months. At the end of
a production cycle, the module is taken back to its support facility. The reactor is thenrefuelled and periodic maintenance is carried out. Major overhauls of the modules arescheduled every 10 years.

165

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

DCNS

Country of origin

FranceReactor type

PWR

Electrical capacity (MW(e))

160

Thermal capacity (MW(th))

530

Expected capacity Factor (%)

91

Design life (years)

60

2Plant footprint (m )N/A for submerged modules

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)15.5oCore inlet/exit temperatures ( C)288/318

Main reactivity control mechanism

Control rods and solid burnable poison

RPV height (m)7.65

RPV diameter (m)3.84

Module weight (metric ton)

25000

Configuration of reactor coolant system

Loop type

Power conversion process

Indirect Rankine Cycle

Fuel type/assembly array

UO2 and Zircaloy cladding/17x17 rods in square assembly

Fuel assembly active length (m)2.15

Number of fuel assemblies

77

Fuel enrichment (%)

4.95

Fuel burnup (GWd/ton)

38

Fuel cycle (months)

38

Cogeneration capability

N/A

Approach to engineered safety systems

Passive

Number of safety trains

2 hydraulic trains and 2 x 3 I&C channels

Refuelling outage (days)

30

Distinguishing features

Transportable NPP, Submerged operation.

Modules per plant

Up to 6 per onshore main control room

Target construction duration (months)

36

Seismic design

Not comparable – subsea conditions

Predicted core damage frequency (per reactor year)< 1 E-7

Design status

Building Partnership

Reactivity Control
A soluble boron free core with enrichment less than 4.95% is used. The absence of boron
reduces the amount of produced waste and improves the safety in comparison with otherreactors (no dilution accident, exclusion of control rod ejection, improved margins on boiling
crisis, low power density, xenon oscillation stability). However, the boron absence leads todifficulties to reach a high burn-up; the core is surrounded by a heavy reflector. Basically, a166

compromise has been adopted between availability, fuel cost and the core size. Furthermore,
to avoid any dependency from the fuel supplier, standard square fuel assemblies are used.
Those ones can be provided by any suppliers and no front end cycle dependency is
introduced in this way.
Reactor Coolant System
The reactor adopts a proven loop-type PWR nuclear steam supply system. The primarycooling mechanism under normal operating condition and shutdown condition is by forced
circulation of coolant. The reactor coolant system (RCS) is designed to ensure adequatecooling of reactor core under all operational states, and during and following all postulated
off normal conditions. Sea water provides a permanent cooling of the reactor core and acts as
the ultimate heat sink.
5. Safety Features
The aim of the design is to make impossible significant radioactive releasing on the
Operating Site when the nuclear unit is anchored on the sub-marine site. The nuclear unit
submersion during the electricity-production phase represents a clear advantage. The safetylevel of the module is greater compared to most of land-based facilities, small or large:•••

Passive systems associated with the infinite heat sink which the sea provides;
Inherent protection against numerous external hazards above the sea (natural or
anthropic) due to the immersion and the water column (aircraft impacts, typhoons andother meteorological effects, etc.);
Reduced intensity of tsunami and earthquake under the sea, in comparison with Land
Based Reactors.

Furthermore, in case of implausible severe accident leading to core meltdown, the defense indepth is based on the In-Vessel Corium retention (I-VCR). The reactor residual thermal
power (the heat decay) is smaller than any large reactors. It allows considering the external
vessel cooling is sufficient to avoid any vessel rupture.
Engineered Safety System Approach and Configuration
The reactor integrates diversity in safety systems and segregates the operating stations toachieve a high level of safety. Water offers a natural protection against many of the possible
external events and guarantees a permanently available heat sink, and an additional barrier tofission products in the case of an accident. The submerged Flexblue reactor has a minimal
environmental footprint. Only the systems required for residual heat removal, control andmonitoring are required in navigation conditions. At the end of its life, the power unit istransported back to a dismantling facility, which results in a quick, easy and full recovery of
the natural site. The use of passive safety systems brings the reactor to a safe and stable state
without external intervention for an indefinite period of time. In particular, the positive
buoyancy of the submerged unit allows extremely efficient de-correlation from the seabed incase of earthquakes. Furthermore, at the depth the unit is fixed, tsunami effects are notcritical. Still, even in postulated extreme situations like large early release of radioactivity inthe water, atmospheric release would be so reduced that it practically excludes any quick
health impact on populations: water quality would have to be watched over, but noevacuation of population would be required. If a safety issue developed with the reactor, it
could be brought to the surface and taken to a shipyard for repair. It could be refuelled in thesame way, and at end of life it would be repatriated to the shipyards for decommissioning.
The emergency power sources (batteries) are sized to ensure the autonomy of the modules up
to 15 days without any external power. The safety of the plant is improved considerably with167

four (4) operating areas in accident conditions: the main control room, the main remote
shutdown station, the local control room and the local remote shutdown station.
Decay Heat Removal System
Decay heat removal is performed by four cooling loops, each one able to remove 50% of
decay heat:••

Two primary chains are connected to the primary circuit: each one includes a heatexchanger immersed in a large safety water tank which is cooled by the ocean throughthe metallic hull.
Two secondary chains are connected to the secondary circuit: each one includes an
emergency condenser directly immersed in seawater.

With the infinite heat sink – seawater – and to the elevation difference of the heat sink withrespect to the heat sources, the four chains operate passively by natural circulation. The longterm safe state of the reactor is a shutdown state where continuous cooling of the reactor coreis achieved by natural circulation.
Emergency Core Cooling System
Protection against loss-of-coolant accidents is ensured by a passive safety injection system
including direct vessel injection lines, core makeup tanks, accumulators and large safety
tanks. All these injection sources inject passively into the vessel In addition, a two-train
automatic depressurization system is connected to the pressurizer. Once these systems haveactuated, a passive recirculation path is in place. Decay heat is removed through the metallic
hull (the containment). In case of an accident, active systems designed for normal/shutdown
core cooling or for controlling coolant inventory are used if alternating-current power isavailable. If not, the passive safety systems described earlier are actuated automatically when
emergency set points are reached. In all accident scenarios, a safe shutdown state is achieved
and maintained for an indefinite period of time without the need for operator action.
Emergency battery power is only required for the opening valves and communication (datatransfer from the module to the remote Control Room). The fourteen (14) days of
autonomous monitoring ability can be extended by reloading the batteries. An emergency
communications system works via radio links in case of loss of cables. Also, an emergency
ultimate system can be run via an acoustic link, when losing both the submarine cables andthe microwave link. This enables to communicate with a ship/submarine and transmits themodule state and enables communications.
Containment System
The reactor containment is bounded by the reactor sector: hull on the sides and reactor sector
walls on the front and on the back. A large share of the metal containment walls are therefore
in direct contact with seawater, which provides very efficient cooling without the need forcontainment spray or cooling heat exchanger. The containment system is also designed tosustain severe accident with core meltdown, although the safety features are designed to
avoid core damage. In this case, the mitigation strategy consists in in-vessel corium retention
assisted by passive ex-vessel core cooling.
6. Plant safety and Operational Performances
The reactor is based on the PWR technology and significantly benefits from a large operating
experience in commercial power plants as well as in the naval environment. The concept
adopts a tele-operation which is a great multiple stakes issue that contributes to the
optimization of the life cycle cost, accident response and management, and ensure a good
availability factor. The module is accessible via a submarine vehicle that connects to access168

hatches, so that maintenance, inspection and operation can be performed on-board while on
the seafloor. The refueling takes place approximately every 3 years. The module is removed
and transported back to a coastal facility, which hosts the spent fuel pool. Major overhaul
occurs every 10 years, i.e., every three fuel-cycles.
7. Instrumentation and Control Systems
The I&C and communications systems are based on different technologies, following thesafety levels in conformity with the IEC standards. Several possible solutions to allocate theclasses of functions to the controllers are considered with different technological andfunctional diversification schemes. I&C system is robust and incorporates experience from
naval industry, especially from nuclear submarines. The I&C system adopts redundancies oftrains, different levels of defence especially on communication systems, technologicaldiversification and the principle of reactor unit I&C autonomy. The I&C system includes adiversified operating station (DOS) enabling commands of safety functions in case oftechnological failure on the nominal station. This diversified system works diversified
solutions from the instrumentation to the man-machine interface (MMI). The main principles
of I&C systems are based on module autonomy towards the shore control facility. The
module autonomy is ensured by the fact that: all the systems supporting the safety functions
are in the module; the reactor can be operated from two different places in the module during
accident conditions; the transition from the initial state to a safe shutdown state when losing
submarine cables is automatic; Local emergency power source. The concept also includes thepossibility of multi-units operations from a single control room in order to run a cluster ofmodules. This aims to reduce the operating crew size considerably. Flexblue employs
redundant main and auxiliary submarine cables that transport electricity as well as
information between the modules and the onshore control centre.
8. Plant Arrangement
The module is composed of a turbine & alternator section, a reactor section, an aft section
and a fore section. These two latter sections accommodate: emergency batteries, a secondary
control room, process auxiliaries, I&C control panels, spares, living areas for a crew andemergency rescue devices. Several units can operate on the same site and hence share thesame support systems.
9. Design and Licensing status
Presentations of the concept have been made to the French safety authority. Technicaldiscussions have been initiated with the French technical safety authority. Preliminary
conclusions have been drawn: these ones lead going ahead the project.

169

KLT-40S (OKBM Afrikantov,
Russian Federation)
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 operationof 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 anddelivered 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.

Figure 1: Reactor System Configuration of KLT-40S(Reproduced courtesy of OKBM Afrikantov)
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, thisFNPP can be used for seawater desalination as well as for autonomous power supply for sea
oil-production platforms.

170

Figure 2: Possible floating power unit configuration(Reproduced courtesy of OKBM Afrikantov)
3. Development Milestones
1998

The first project to build a floating nuclear power plant was established2002

The environmental impact assessment for KLT-40S reactor was approved by the RussianFederation Ministry of Natural Resources

2006

After several delays the project was revived by Minatom (Russian Federation Ministry of Nuclear
Energy)2012

Pevek (Chukotka Autonomous Region) was selected as the site for the installation of small
modular NPPs integrated with floating platforms with the KLT-40S reactor. JSC “Baltiysky
Zavod” (Saint Petersburg) undertook charge of construction, installation, testing andcommissioning the first floating plant named Akademik Lomonosov (2 units with the output of 35
MW(e))
Completion of construction and testing of the floating power unit at the Baltic shipyard

20172019-2020

4.

Commercial startup

General Design Description

Design Philosophy
KLT-40S is the reactor for FNPP Akademik Lomonosov, intended for reliable power and heat
supply to isolated consumers in remote areas without centralized power supply and where
expensive imported fuel is used.
Nuclear Steam Supply System
The steam lines while exiting from the SGs are routed through containment to a set of steam
inlet valves, and finally into the turbine building for electricity conversion. Cogenerationequipment could be modified into the medium-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 theadvantages foreseen by the FNPP under construction is long term independent operation inremote regions with decentralized power supply. The design requires refueling of reactor after
every 3–4 years of operation. Refueling is performed 13 days after reactor shutdown when
the levels of residual heat releases from spent FAs have reached the required level. The spent171

fuel will be delivered to the importing country. The spent nuclear fuel will then be stored on
board the FNPP. No special maintenance or refueling ships are necessary. Single fuel loading
is done in order to provide maximum operation period between refueling. The fuel is loadedin the core all at once with all fuel assemblies being replaced at the same time. Besides, lowenriched U235 with enrichment below 20% is used to meet non-proliferation requirements.
Reactivity Control
The control rod drive mechanism (CRDM) is electrically driven and releases control andemergency control rods into the core in case of station black out (SBO). The speed of safetyrods driven by electric motor, in the case of emergency is 2 mm/s. The average speed ofsafety 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 isused only in the emergency 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 connected with short nozzles. The reactor has a four loop system with forced andnatural circulation, a pressurized primary circuit with canned motor pumps and leak tightbellow type valves, a once-through coiled SG and passive safety systems. KLT-40S thermalhydraulic connections comprising external pressurizer, accumulators, and separation heat exchanger are in proximity of the reactor systems. The pressurizer is not an integral part of thereactor systems and in this design it is formed by one or more separate tanks, designed toaccommodate changes in coolant volume, especially severe during reactor start-up. The core
is cooled by coolant 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 thecold 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).
MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

Afrikantov OKBMCountry of origin

Russian Federation

Reactor type

PWR

Electric power (MW(e))

35Thermal power (MW(th))

150

Expected Capacity Factor (%)

60 to 70

Design Life (years)

40

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)12.7oCore inlet/outlet temperatures ( C)280/316

Steam capacity (t/h)

240

172

Overheated steam pressure at SG outlet (MPa)

3.82

o

Overheated steam temperature at SG outlet ( C)290

Main Reactivity Control Mechanism

Control Rod Drive Mechanism

RPV Height (m)4.8

RPV Diameter (m)2.0

SGU weight (without primary and secondary coolant) (t)211

Configuration of the Reactor Coolant System

Compact Loop TypePower Conversion Process

Indirect Rankine Cycle

Fuel Type/Assembly Array

UO2 pellet in silumin matrix

Fuel Assembly Active Length (m)1.2

Number of Fuel Assemblies

121

Fuel Enrichment (%)

18.6

Fuel Burnup (GWd/ton)

45.4

Fuel Cycle (months)

30-36

Cogeneration Capability

Possible

Approach to Engineered Safety Systems

Active (Partially passive)

Number of safety trains

2

Refueling Outage (days)

30 – 36Distinguishing features
Modules per Plant

Floating power unit for cogeneration of heat and electricity;
Onsite refuelling not required; Spent fuel take back to the
supplier
2

Generator terminal power (MW(e))

2 x 38.5 (gross)Maximum thermal power (Gcal/h)

2 x 73

Coastal area (ha)1.5

Water area (ha)6.0

Number of operating personnel (persons)< 60

Target Construction Duration (months):

48 – 60 (including manufacturing of components)

Seismic design

9 point on the MSK scale

Predicted core damage frequency (event/reactor year)

0.5E-7

Length of the ship (m)140Width of the ship (m)30Hull height (m)10Full load weight (ton)

21000

Cogeneration plant service period (year)

40

Design Status

Under construction, planned commercial start 2019-2020

5. Safety Features
The KLT-40S is designed with proven safety aspects such as a compact structure of the SG
unit with short nozzles connecting the main equipment, primary circuit pipelines with smaller
diameter, and with proven reactor emergency shutdown actuators based on different
operation principles, emergency heat removal systems connected to the primary andsecondary circuits. Additional barriers are provided to prevent the release of 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 coolingtrain through primary circuit purification system’s heat exchanger thermally coupled with a173

’third’ independent circuit exchanging heat energy with ambient sea or lake water, active
cooling train through the SGs heat exchangers with decay heat removal accomplishedthrough the condenser which in turn is cooled down by ambient sea or lake 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 andpassive safety systems are to perform the reactor emergency shutdown, emergency heatremoval 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 onprotection 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 andetc. Passive cooling channels with water tanks and in-built heat exchangers ensure reliablecooling to 24 hours. Actuation of the system is also performed by special devices with
passive actuation principle – hydraulically operated pneumatic valves.
Engineered Safety System Approach and Configuration
The active components of the protection system are scram actuators for six groups of thecontrol rods.
Decay Heat Removal System
The decay heat removal system is intended to remove core residual heat upon actuation ofreactor emergency protection in case of abnormal operation including accidents, as well as toremove residual heat at normal RP decommissioning. The decay heat removal system
includes the following cooling channels:•••

Two secondary passive cooling channels via steam generators;
One active secondary cooling channel via steam generators;
One active cooling channel via the primary/third heat exchanger.

Reactor core heat is removed by two passive system channels at de-energization for morethan a day.
Emergency Core Cooling System
The Emergency Core Cooling System (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;• Hydraulic testing of the primary circuit and associated systems, secondary and third
loop sections disconnected at inter-circuit leaks and designed for primary pressure.
The ECCS includes the following subsystems:• High-pressure ECCS subsystem with makeup (low-capacity high-pressure) pumps;• High-pressure ECCS subsystem with hydraulic accumulators;• Low-pressure ECCS subsystem with recirculation pumps.

174

PASSIVE SYSTEM OF
EMERGENCY PRESSURE
DECREASE IN THE
CONTAINMENT (CONDENSATESYSTEM)PASSIVE EMERGENCY
CORE COOLINGSYSTEM (HYDRAULICACCUMULATORS)ACTIVE EMERGENCY
CORE COOLING SYSTEMPASSIVE
EMERGENCY
SHUTDOWNCOOLINGSYSTEM

Purification
FROM STP (Standby andemergency feed water pumps

ACTIVE SYSTEM OF LIQUID
ABSORBER INJECTION

ACTIVE SYSTEM
OF EMERGENCYSHUTDOWNCOOLING
THROUGH
PROCESSCONDENSERSPASSIVE SYSTEM OF
EMERGENCY PRESSURE
DECREASE IN THE
CONTAINMENT(BUBBLING SYSTEM)ACTIVE EMERGENCY
CORE COOLINGSYSTEM
RECIRCULATIONSYSTEM PUMPSSYSTEM OF REACTORCAISSON FILLING
WITH WATER

Figure 3: KLT-40S Emergency Core Cooling System flow diagram(Reproduced courtesy of OKBM Afrikantov)
Containment system
The containment for the KLT-40S is configured for FNPP applications and is made of steel
shell designed to sustain mild pressurization, while the reactor systems are positioned inside areinforced “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 portionsof 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 forrefueling or maintenance operations.
6. Plant Safety and Operational Performance
The KLT-40S NPP ensures electricity and heat generation within the power range of 10% to100% for a continuous operation of 26000 h. The NPP is designed for manoeuvring speed of
up to 0.1%/s. As a countermeasure against the external impact, the NPP is fitted with both
ground safety and floating physical protection means. Structures are designed to be placed in
the Arctic zone at the depth of 2 m at freezing temperatures. The Floating Power Unit (FPU)
and NPP buildings are designed to withstand the crash of an aircraft of 10 t. Based on theanalysis of emergency situations, the radiation-ecological impact upon the personnel andenvironment at normal operation, abnormal operation, including design-basis accidents does
not result in exceedance of established dose rates for the personnel and population, thecontent of radioactive substances in the environment, and it is limited at beyond-design-basis
accidents.
7. Electric Power System
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
175

diesel generators of 992 kW each. The emergency electric system supplies electricity tosafety system loads in all operation modes, including loss of operating and back-up electric
power sources. The FPU has independent emergency electric systems for each reactor plant.
Each emergency electric system has two channels with an emergency diesel generator of 200
kW.
8. Plant Arrangement
The coastline line of the Floating Cogeneration Nuclear power Plant (FCNPP) has thefollowing auxiliary facilities:••

Complex engineering building with equipment to distribute and transmit electricity toloads and to prepare and transfer heating water to loads;
Auxiliary buildings, including: two hot water storage tanks; partially in ground tankwith slime water; wet storage bunker; two cooling towers; access control point; site
enclosure; lighting towers.

The coastal line of the FCNPP does not provide for handling nuclear materials and radiation
hazardous media.
Waterside structures

Floating power unit

Thermal power
Transmission device
Electricity
Transmission device
Enclosure

Conveyor bridge

10 kW cable in the
trenchHoses for rigid docking

Access control
point

Astern pierHeating waterPreparation point 10 kW
cable in the trench

Figure 4: FCNPP Arrangement (Reproduced courtesy of OKBM Afrikantov)
Reactor Building
The general cross-section view of the FPU is given in Figure 5. The FPU is a flush deck nonself-propelled rack-mounted vessel with hull and multi-layer deckhouse. The medium portionof the FPU has a reactor compartment and nuclear fuel handling compartment. A
turbogenerator compartment and electro-technical compartment are arranged in the ship’s
head with respect to the reactor compartment, auxiliary installations compartment andaccommodations are arranged in astern. Each reactor 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. On-board the FPU, storages forspent cores and means are arranged that ensure reactor reloading.
Control Building
The KLT-40S reactor is controlled using the operator’s automated workstation throughrespective control panel located in the central control room. In case it is impossible to carryout control from the central control room, information on the reactor status is obtained andsafety systems are activated to make reactors subcritical and control reactor plant coolingusing emergency cooling control panels located outside the central control room.

176

StorageRP

Accommodations

STP

Figure 5: General cross-section view of the FPU (Reproduced courtesy of OKBM Afrikantov)
Turbine Generator Building
The steam turbine plant (STP) is intended to convert the thermal power from steam obtainedin 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 is connected 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 through the Akademik Lomonosov FNPP. Overall, the KLT-40S is a modified
version of the commercial KLT-40-type propulsion plants employed by the Russian
icebreaker fleet. The environmental impact assessment for KLT-40S reactor systems wasapproved 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 (Russia’s nuclear regulator). The keel of the first FNPP carrying
the KLT-40S, the Akademik Lomonosov in the Chukotka region, was laid in 2007. The
Akademik Lomonosov is to be completed by the end of 2016 and expected electricity
production in 2018 – 2019.
10. Plant Economics
It is expected that the economic competitiveness of the KLT-40S reactor units would bedetermined by the following advantages:•••

FNPPs can be serially manufactured in shipyards and delivered to a customer fully
assembled and ready for operation;
The reactor is based on a well-proven design; the layout of the reactor is compact andmodular, reduced amount of capital needed to prepare FNPP’s docking area ascompared to a site for a land-based NPP;
Reduction in construction period, and operating costs due to need for less on-site
employees.

Specific capital costs for the FPU are estimated as 5,300 $/kW(e) according to rate of 2016.
Coastal and waterside structures of the FPU depend on the site location. The production costof electricity generated by the KLT-40S FPU reactor plant is estimated as 8.8 cent/kW
according to rate of Quarter I, 2016 at a discount rate of 7% (less than 10 cent/kWh).

177

RITM-200 (OKBM Afrikantov,
Russian Federation)
1.

Introduction

The RITM-200 is an integral pressurised water reactor (PWR) for multipurpose nuclear
icebreaker, floating or land-based NPP producing an electrical output of 50 MW, designed by
the OKBM Afrikantov. Initially, the RITM-200 was developed specifically for nuclear
icebreaker ships powered by two reactor modules. Later on, it was designed for nonpropelled floating nuclear power plant (FNPP). The RITM-200 can be further modified to
produce heat for desalination plants and various industrial applications, e.g, marine drilling
platforms. The new generation icebreaker ship using RITM-200 is designed with double-hulland an increased icebreaking capability. The RITM-200 will basically replace the KLT-40
series for both nuclear icebreaker ship and land based power plant. RITM-200 will require
refuelling every 7 years in its 40-year design life.

Figure 1: Reactor System Configuration of RITM-200(Reproduced courtesy of OKBM Afrikantov)
2. Target Application
The RITM-200 is designed to provide the shaft power on a nuclear icebreaker. The reactor
can also be used for FNPP for cogenerating heat and electricity, power and desalination
complexes, and offshore drilling rigs.

178

3. Development Milestones
2009

Detailed design of the reactor plant finished

2012

2016

The detailed design documentation developed for the main equipment, i.e. reactor, circulation
pump. CRDMs, primary circuit filter, heat removal system HX, safety devices and valves.
Fabrication of components and equipment started.
Complete RPV delivery

2017

The first RITM-200 icebreaker will be in operation

2020

Two serial RITM-200 icebreakers will be commissioned

4. General Design Description
Design Philosophy
An integrated approach was adopted to determine the main parameters of the primary system,
selection of equipment and layout, determining the optimal inventory and parameters of thesafety systems for the RITM-200. Inherent safety characteristic is ensured by high thermal
storage capacity, natural circulation capacity to cool the primary system, primary systempipelines with minimum length, nozzles equipped with leak stoppers, and large volume ofprimary coolant for enhanced time margin the case of loss of coolant accidents. The reductionof neutron fluence in the vessel enables the increase in radiation margin.
Nuclear Steam Supply System
The RITM-200 has an integrated nuclear steam supply system (NSSS) in which 4 steam
generators are installed inside the RPV.
Reactor CoreThe RITM-200 adopts a low enriched cassette core similar to that of the KLT-40S that
ensures long time operation without refuelling and meets international non-proliferation
requirements. The core houses 199 fuel assemblies with fuel enrichment up to 20%. The
design also gives lower fluence on the reactor vessel. The core has the assigned service life of
7.0 TWh that enables it to achieve 65% capacity factor.
Reactivity Control
Control rod drive mechanism (CRDM) is used for reactivity control. A bank of emergency
protection (EP) drive mechanisms is designed for fast reactor shut down and to maintain it in
the subcritical condition in accidents. A bank of compensation group (CG) drive mechanisms
is intended to compensate for the excessive reactivity at start up, power operation and reactor
trip. The CRDM of the RITM-200 is made based on the drives used in the KLT-40S.
Reactor Coolant System
The reactor is designed as an integral vessel with the main circulation pumps located inseparate external hydraulic chambers with side horizontal sockets for steam generator (SG)cassette nozzles. Each of the four SGs have 3 rectangular cassettes, while the four main
circulation pumps are installed in the colds leg of the primary circulation path and separated
into four independent loops. The SGs produce steam of 295°C at 3.82 MPa flowing at 248
t/h. The conventional RCPs are used. The pump is vane, single step, and has a sealed
asynchronous electric motor. The motor has one winding. The rotation frequency is adjusted
using conversion of the power current frequency.

179

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

OKBM Afrikantov

Country of origin

Russian Federation

Reactor type

Integral PWR

Electrical capacity (MW(e))

50Thermal capacity (MW(th))

175

Expected capacity factor (%)

65

Continuous operation period (hr)

26000

Design life (years)

40

2Plant footprint (m )N/ACoolant/moderator

Light water

Primary circulation

Forced circulation by pumps

System pressure(MPa)15.7oCore inlet/exit temperatures ( C)

277 / 313

Core flow rate (ton/hr)

3,250

Steam generating capacity (ton/hr)

248

oSteam temperature ( C)295

Pressure (MPa abs)

3.82
2

Neutron fluence at the end of life time (neutron/cm )5.2 x 1019

Main reactivity control mechanism

Control Rod Drive Mechanism (CRDM)

RPV height (m)8.5

RPV diameter (m)3.3Module weight (metric ton)

1100

Configuration of reactor coolant system

Integral PWRPower conversion process

Indirect Rankine cycle

Fuel type/assembly array

UO2 pellet/hexagonal

Fuel assembly active length (m)1.2 – 1.65

Number of fuel assemblies

199

Fuel enrichment (%)< 20

Fuel burnup (GWd/ton)

68.4 / 51.2

Fuel cycle (months)

54 – 84

Cogeneration capability

Yes

Approach to engineered safety systems

Combined active and passive

Number of safety trains

2

Refuelling outage (days)

40

Distinguishing features

Developed for universal nuclear ice breakers

Modules per plant2

Target construction duration (months)

48

Seismic design (g)3gCore damage frequency (per reactor-year)

0.9E-6

Design Status

Components under construction, ready for installation

180

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) houses the reactor core and internals and four (4) steam
generators. Four (4) vertical reactor coolant pumps are directly connected onto the RPV using
pipe stubs.
Steam Generator
The RITM-200 uses once through (straight tube) SGs, of which the specific steam generating
capacity is more than two times higher than the capacity of the common coil-type 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 power engineering. It is characterized by a simple design, which increases reliability,
compactness, and no electric power required. The compensation system is divided into twoparallel independent groups to reduce the restrictor diameter in the compensatory nozzles ofthe steam generating unit and to decrease a coolant leak rate in primary-pipe-break accidents.
It makes possible to use one of pressurizers as a hydraulic accumulator, increasing reactor
plant reliability considerably in potential loss-of-coolant accidents.
5. Safety Features
The safety concept of the RITM-200 is based on the defence-in-depth principle combinedwith the plant self-protection and use of passive systems. Properties of intrinsic safety areintended for automatic control of power density and reactor auto-shutdown, limitation ofprimary coolant pressure and temperature, heating rate, primary circuit depressurization
scope and outflow rate, fuel damage scope, maintaining of reactor vessel integrity in severe
accidents and form the image of a “passive reactor”, resistant for possible disturbances.
RITM-200 optimally combines passive and active safety systems to cope with abnormal
operating occurrences and design basis accidents.
••••

Passive pressure reduction and cooling down systems have been included (system
reliability is confirmed by bench testing);
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 theplant during SG and RCP failures.

The exposure dose for the crew in normal operation and design basis accidents does notexceed 0.01% of the natural radiation limit. The public exposure dose in beyond DBA withcore severely damaged, does not exceed the values which require protective measures.
Seawater activity due to operation is equal to 0.1 Bq/l, which is 100 times less than thespecified value for drinking water.

181

Figure 2: Emergency Core Cooling System of RITM-200(Reproduced courtesy of OKBM Afrikantov)

182

Engineered Safety System Approach and Configuration
The RITM-200 high safety level is achieved both by inherent self-protection properties and a
combination of passive and active safety systems. Moreover, redundancy of safety systemequipment and channels and their functional and/or physical separation are provided toensure high reliability indices. 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’s medium
pressure to ensure reactor trip and initiate the safety systems. The CRDM is electricallydriven that will insert emergency control rods into the core in case of station black out (SBO).
The speed of safety rods in the case of emergency when driven by electric motor is 2 mm/s.
The average speed of safety rods if driven by gravity is 30 to 130 mm/s.
Decay heat removal system
The decay heat removal system (DHRS) combines several independent channels connected todifferent SGs. Each channel contains a DHRS tank; a heat exchanger located in the DHRS
tank; a water storage tank, and air cooled heat exchanger. The system uses two types of heatexchangers (HXs): water-cooled HX similar to those used in KLT-40S and air-cooled HX. At
the initial stage, heat is removed by the two HXs in which the air goes directly to theatmosphere and the water goes to the DHRS tank to ensure the desired optimum heatremoval. After the water in the DHRS tank evaporates, the air-cooled HX would still operate
and provide cooling to the core for unlimited period. This combination enables a minimized
overall dimension of the HXs and water storage tank.
Emergency core cooling system
The Emergency core cooling system (ECCS) is based on active and passive operating
principles, a two-channel design with redundancy of active elements in each channel. Eachchannel comprises hydraulic accumulators supplying water to the SG passively by gaspressure, and makeup pumps supplying water to the SG when power supply is available.one
group of pressurizers is used as hydraulic accumulator when the other group of pressurizers isdepressurized.
Containment system
The RITM-200 module is installed inside containment with a volume of 6m×6m×15.5m. The
integrated NSSS configuration significantly reduces the reactor dimension inside thecontainment. The mass of the containment with reactor module is 1100 ton. The containmentpressure reduction system is passive-driven and provides localization of steam-water mixture
during LOCA in the containment of the damaged RITM-200 module, by using self-actuating
devices like hydraulically operated pneumatic valves (at increase of pressure in containment)
and pressure-actuated contact breakers.
6. Plant safety and Operational Performances
The assigned service life of the plant for replaceable equipment is 20 years with a continuous
operation period of 26000 h.
7. Instrumentation and Control systems
An automated control system is provided in the RITM-200 nuclear power plant to monitor
and control plant processes. This system possesses necessary redundancy with regard tosafety function fulfilment and allows both automated and remote (operator’s) control of thepower plant.

183

8. Plant Arrangement
The basic architectural design for the land-based power unit is to place the reactor , including
its servicing systems, spent fuel pool, and auxiliaries, in a double protective containment
resistant to the air-crash. The designers also claim that the overall size of the module allows
transport of the reactor by rail.
Reactor building
RITM-200 plant with service equipment is arranged in a building providing protection
against external impacts. The building structure is also provided with biological shielding tolimit impact on the environment and population during reactor operation.
Control building
Station operation is controlled from the main control room by plant operators . Information
control and representation means possess necessary redundancy to provide station operation
monitoring and control in emergency and failure conditions.
Balance of plant:
The RITM-200 design can be configured for land-based stationary applications wherein thereactor system – nuclear island coupled to a turbine island and ancillary buildings for spent
fuel storage, water treatment, maintenance, and switchyard connections with configurations
similar to conventional large LWRs, are housed within a relatively small footprint.
Turbine Generator building
Each RITM-200 reactor can be thermally coupled with one or multiple turbinegenerator 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 heatexchanger of a district heating circuit.
Electric power systems
The electric power system consists of power generation system and power distribution
system. The structure of this system depends on a specific station location andCustomer’s needs. The house load power supply system for station consumers is also
incorporated which provides power supply for station consumers under normal andemergency conditions.
Floating nuclear station based on the RITM-200 reactor plant
The RITM-200 reactor plant also could be a base to create a floating nuclear power
station. Moreover, it is possible to operate this station without refueling at its site. This
fact will make it more commercially attractive, and availability of near term prototype –the power plant of the universal nuclear icebreaker – will reduce time and cost of thisplant.
9. Design and Licensing Status
The RITM-200 design was developed in conformity with Russian laws, norms and rules formarine nuclear power plants and safety principles developed by the world community and
IAEA recommendations. An order for the first commercial FNPP with RITM-200 is expected
to be placed in the near future, while the first icebreaker adopting this technology is expected
to be commissioned in the next few years. The RITM-200 adopts optimal combination ofpassive and active safety systems. The components for the two modules of RITM-200 are
being manufactured for the first multipurpose icebreaker targeting potential completedelivery in 2016. Serial deliveries of reactor modules for two consequent nuclear ice-breakers
184will be in 2017 and 2018. RITM-200 has been chosen for the next generation of RussianFederation ice-breakers and accordingly costs for the design will decrease as reactors are built
and installed.
10. Plant Economics
The average specific cost of construction for the land-based station with the RITM-200 plant
will be approximately 6000$/kW, while the electricity unit cost is approximately
0.095$/kW·h. The average specific cost of construction for the floating station with the
RITM-200 plant will be approximately 3900$/ kW, while the electricity unit cost isapproximately 0.062$/ kW·h. The shown cost characteristics do not consider coastal
facilities and infrastructure of the station, because they depend considerably on a site
(geophysical data and climatological conditions, valid local regulatory documents, social
economic and infrastructure conditions, peculiarities of a selected operation model and otherfactors).

185

VBER-300 (OKBM Afrikantov, RussianFederation)
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 evolutionof modular marine propulsion reactors. An increase in thermal power causes an increase inmass and overall dimensions; however, the reactor basic design is similar to that of marine
propulsion reactors. The VBER-300 design was developed based on the lessons learned fromthe design, safety and operating experience for VVER reactors. VBER-300 adopts provennuclear ship building technologies and operating experience that in turn contribute toenhancement of operational safety and reduction in production costs. VBER-300 can be
configured as a multi-module plant on request of the customer. VBER-300 design features areavailability for both land-based and transportable FNPPs, a variety 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.

Figure 1: VBER-300 Reactor System Configuration(Reproduced courtesy of Afrikantov OKBM)
2. Target Application
The VBER-300 nuclear plants are intended to supply thermal and electric power to remoteareas 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 VBER-300 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 300000. 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 for process heat applications.

186

3. Development Milestones
2001

Design activities to develop VBER reactors started

2002

Technical and commercial proposal for the two-unit VBER NPP

2004

2007–08

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, automatedprocess control system, and heat-generating plant; feasibility, economy, and investment
studies of the VBER-300 RP NPP for the Mangistau Region, Kazakhstan
Development of the 100–600 MW VBER plant

2008–11

R&D for the VBER-460/600 NPP design

2011–12Development 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
Development of the conceptual design for the two-unit VBER-600/4 NPP

2006
2007–09

2012–2015

4. General Design Description
VBER-300 technical characteristics are modular design, once-through steam generators withtitanium tubes, canned main circulation pumps, low density core, and optimal combination ofactive and passive safety systems. VBER-300 has some specific features, such as a double
containment with low possibility of radioactive release under all accident scenarios, anultimate heat sink by air cooling at minimum requirement for water cooling, a reduced
nuclear steam supply system (NSSS), a simplified balance of plant, and suitability for bothland-based stations and transportable FNPPs.
Design Philosophy
VBER-300 design using ship-based modular configuration enhances the safety philosophy
through proven marine modular technologies. The reactor design has no pipelines in the
primary circulation circuit. All vessels of main equipment such as reactor vessel, steam
generators , and main coolant pump(MCP) are welded to each other to make a single reactor
unit as shown in Figure.1. The VBER-300 reactor unit incorporates the reactor and foursteam generator – MCP two-vessel units. The compact VBER-300 reactor system comprises
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 be arranged to operate individually or as multi-module plant, increasing thepower output by means of scaling up the equipment and with the same reactor system
configuration.
Reactivity Control
In the VBER-300 reactor, 61 control rods in combination with fuel elements mixed withburnable poison materials provide safe and reliable reactivity control during both normal andtransient operating period. Control rods are operated through high-performanceelectromechanical control rod drive mechanisms (CRDMs). The control rods elements aredesigned to maintain the core subcritical even if the most reactive assembly fails (i.e. stuckrod/assembly event). To compensate for the fuel burnup reactivity margin, fuel rods withgadolinium burnable poison contained in uranium dioxide pellets are distributed across each
FA with configurations similar to those used in VVER-1000 reactors. Boric acid is also
dissolved and maintained at controlled concentrations within the primary coolant system toensure optimum core power distribution.

187

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Afrikantov OKBMCountry of origin

Russian Federation

Reactor type

Integrated PWR

Electric power (MW(e))

325

Thermal power (MW(th))

917

Expected capacity factor (%)

92

Design life (years)

60

Plant area (ha)30 (two-unit nuclear cogeneration plant)

Coolant/moderator

Light water

Primary circulation

Forced circulation

System pressure (MPa)16.3oCore inlet/exit temperatures ( C)

292 / 328

Main reactivity control mechanism

Control Rod Drive Mechanism and soluble boron

RPV height (m)9.3

RPV diameter (m)3.9

RPV weight (metric ton)

200

Height of the reactor unit, m

15.3

Circumscribed diameter of the reactor unit, m

11.2

Mass of the reactor unit (w/t water), t

1,435

Configuration of reactor coolant system

Integrated

Power conversion process

Indirect Rankine Cycle

Fuel type/assembly array

UO2 pellet/hexagonal

Fuel assembly active length (m)3.5

Number of fuel assemblies

85

Fuel enrichment (%)

4.95

Fuel burnup (GWd/ton)

50

Fuel cycle (months)

72

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (active and PASSIVE) system

Number of safety trains

2

Refuelling outage (days)

Maximum 24 days to refuel the core and carry out periodic
maintenanceMaximum 16 days only to refuel the core

Specific features
Modules per plant

Power source for transportable FNPPs, a variety of cogeneration
i ll
d i
h
d ffi i
1 iTarget construction duration (months)

40

Seismic design (g)0.25Core damage frequency (per reactor-year)

1E–6

Design status

Licensing stage

188

Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) consists of the reactor core and internals with an overall
height of 9.3m and a diameter of 3.9m. The VBER-300 design provides a special system of
emergency vessel cooling to solve the problem of retaining the melt inside the reactor vesselin severe accidents. The core melt retention is facilitated by the low power density, relatively
low level of residual heat, no penetrations in the reactor vessel bottom and smooth outersurfaces 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 and operational safety of the MCPs are enhanced due to the usageof 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 componentsof the primary loop are directly connected to the RPV, except for the pressurizer. The MCPs
are centrifugal single-stage canned pumps with impellors.
Steam Generator
The SGs are once-through coil modules with the secondary coolant flowing inside the tubes.
The feedwater is pumped through an inlet in the SG head, circulates within the SG tubes andexits through the SG outlet as a superheated 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 the pressurizer, 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 thecold 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 headin the steam region has two safety valves that protect the primary circuit against overpressure
in case of accidents with loss of decay heat removal.
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).

189

Figure 2: VBER-300 Emergency Core Cooling System(Reproduced courtesy of Afrikantov OKBM)
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 andenhancement of safety against external events (including acts of terrorism); and mitigation ofsevere accident consequences. The RPV and connecting piping that usually form the primarypressure boundary represent an additional physical barrier. The leak-tight carbon steel
containment and protective enclosure with filtration forms the ultimate barrier separating thereactor system from the environment. For all cogeneration applications, the separation of heatexchangers represents a physical barrier to prevent radioactive release. For the VBER-300
district heating configuration as shown in Figure. 3, a physical barrier is formed by the
independent thermal-hydraulic loop comprising the district heating pump, associated piping,and radiators.
Decay Heat Removal System
The decay heat removal system (DHRS) consists of two passive heat exchangers and aprocess condenser. Passive safety features are intended to arrange recirculation in the core forthe removal of decay heat in the course of scheduled maintenance, refuelling or underemergency conditions. Passive emergency shutdown cooling system operates using natural
circulation of coolant in all heat transport circuits with stored water tanks, 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 core cooling 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 accumulators operate in twostages. The VBER-300 emergency shutdown system consists of the control rod drive
mechanisms, 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
190

cooling channels with water tanks and in-built heat exchangers, ensure reliable cooling 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 is done using compressed gas. Containment depressurization
systems prevent containment damage and reduce radioactive release in design basis accidents(DBA) and beyond DBAs. A small and medium loss of coolant accidents (LOCA) areprevented 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 hydraulically connected 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 naturaland man-caused impacts.
6. Plant Safety and Operational Performances
The VBER-300 safety concept is based on the state-of-the-art defense-in-depth principles
aimed at prevention of accidents and mitigation of accident consequences, including
consequences of severe accidents, mainly due to well-developed self-protection properties ofthe modular PWR and wide application of passive and active safety systems. With themodular configuration, the VBER-300 has increased resistance to impact loads in case of anearthquake and an 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
VBER-300 reactor unit has a two-times safety margin (maximum seismic stress of the mostloaded vessel unit is 150 MPa maximum at an allowed stress of 370 MPa). Analysis of the
20 t aircraft crash on the VBER-300 reactor compartment showed that overload upon theattachment fittings of the reactor unit is less than seismic loads. It is considered that coremelting accidents for the VBER-300 core are postulated. In the course of the severe accident,
the reactor cavity is filled with water from the emergency reactor vessel cooling system
ensuring reliable heat removal from the external surface of the bottom and lower portion of
the vessel. Retention of satisfactory mechanical properties and load-carrying capacity of thevessel ensures retention of the melted core inside the reactor. 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 the perimeter of the industrial site. The calculated radius of protection
measures planning for population is 1 km.
7. Plant 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.

191

Separation HXs

Figure 3: VBER-300 electric and district heating application(Reproduced courtesy of Afrikantov OKBM)
VBER-300 as a Transportable Floating Nuclear Power Plant
VBER-300 is a non-propelled transportable FNPP designed according to the Russia’s Sea
Navigation Register. The transportable FNPP is located on a platform consisting of threepontoons (one central and two peripheral ones). The independent units are in-line located on
the central pontoon. Each unit consists of a reactor compartment, power plant control board
and electrical equipment compartment; areas for the plant refuelling and repair are also
provided. A steel protective shell (containment system) houses main equipment of the reactor
module with related service systems. Fuel storage is located on the central pontoon between
the units. Turbine generators and their related equipment and systems are located in stern andfore sections of the central pontoon. Electrical equipment for power transformation,
distribution and supply (up to 220 kV) to coastal facilities and that for power supply to theplant in-house loads are located on the left board pontoon. A composite steel-iron concrete
vessel of the power unit eliminates the need for scheduled docking during the plant lifetime.
A period between overhauls (20 years) is defined by the key reactor plant equipment life thatis about 15000 h. The VBER-300 FNPP operating life is 60 years.

192

Reactor Building
The inner steel shell is a leak-tight cylindrical enclosure 30 m in diameter that is covered withthe semispherical dome 15.0 m in radius and that has an elliptical bottom. The height of theleak-tight enclosure is 47.0 m. The steel shell is designed for parameters of the maximum
DBA with the excess pressure of 0.4 MPa and the temperature of 150 C. The outer protective
enclosure is made of one-piece reinforced concrete without preliminary tensioning of thesteel and consists of a cylindrical portion of the semispherical dome. Building structures ofthe outer protective enclosure are designed for external 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 thereactor 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—is 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 heatexchanger of a district heating circuit. A VBER-300 nuclear station can operate as a
NPP with a condensing 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 technical design.
9. Plant Economics
Calculations of construction costs and electric power generation costs performed in 2006–
2009 for VBER-300 nuclear units showed that these can compete with carbon electric powersources for isolated regions that are not included into large power systems and for regionsthat strongly depend on the cost of fossil fuel purchase and transportation.

193

ABV-6E (OKBM Afrikantov,
Russian Federation)
1. Introduction
The ABV-6E is reactor plant (RP) as a part of nuclear power system (NPS) that produces
14 MW(th) 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 wasdeveloped using the operating experience of WWER reactors and recent achievements in thefield of nuclear power plant (NPP) safety. The main objective of the project is to develop
small, shop fabricated, multipurpose transportable NPP for safe operation during 10 – 12years without refuelling at the berthing platform or on the coast. Plant maintenance andrepair, refuelling and nuclear waste removal will be carried out at dedicated facilities.

Figure 1: Reactor System Configuration of ABV-6E(Reproduced courtesy of OKBM Afrikantov)
2. Target Application
The ABV-6E RP is intended as a multi-purpose RP. The RP is designed with the capability of
powering a floating power unit (FPU) as a part of floating nuclear power plant (FNPP) with amaximum 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 heatand power cogeneration or can be used for other applications. Besides, a land-based
configuration of the plant is also applicable.

1943. Development Milestones
1993

The final design was developed for the prototype reactor plant and Volnolom floating NPP

2006feasibility 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 the contract
with Minpromtorg (Russian Federation Ministry of Industry and Trade)

2007
2014

4. General Design Description
The basic design approach for the ABV-6E is optimization of the capital and operationalcosts. The ABV-6E adopts design methodologies for minimizing mass and overall
dimensions, feasibility to deliver Reactor Pressure Vessel (RPV) by railway or watertransport with minimal installation activities at the construction site. Therefore, single-reactor
power units are the most preferable designs. From the economic point of view it reduces thestartup capital costs, construction and payback periods. For the FPUs it provides significant
reduction of displacement and feasibility to locate them in the sea shallow water areas and in
the rivers. The ABV-6E is particularly applicable for modular-transportable power unit
(MTPU) in sea shallow water areas. The land based MTPU with ABV-6E is fabricated aslarge, ready-made modules. Due to modularization of the main equipment into large–size
prefabricated modules the scope of construction and installation activities to be performed at
the installation site would be significantly reduced. The delivered modules are installed on
the prepared foundation, interconnected by appropriate pipelines, and then easy-to-install
modular structures are assembled, providing protection of the power unit against the external
factors. Structural optimization of main equipment allows for delivery of the power units atthe base site through water with minimum technological issues.

Figure 2: The conceptual view of the MTPU with ABV-6E(Reproduced courtesy of OKBM Afrikantov)

195

Figure 3: The conceptual view of the FPU with ABV-6E(Reproduced courtesy of OKBM Afrikantov)
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% by weight (maximum
enrichment 19.7%).

Nuclear Power System
NPS consists of 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 and the control rod drive mechanism islocated above the shield outside the vessel.
Reactor Core
The core comprises 121 hexagonal fuel assemblies (FA) of cassette type with active partheight of 900 mm, similar to the FAs in KLT-40S. Cermet fuel is used with maximum 19.7%
enriched uranium-235. Special stainless steel is used as fuel cladding.
Reactivity Control
Reactivity control without boron solution in the primary coolant and compensation ofreactivity changes in power operation is achieved by mechanical control and protection
system (CPS). These inherent safety features ensure automatic power regulation in a steadystate operation, self-limiting power rise in case of positive reactivity insertions, automatic
control of the reactor power and primary coolant pressure and temperature in transients, aswell as the emergency shutdown of the reactor core including the cases with a blackout andRPV flip-over (with account of the time that the vessel flip-over process takes).

196

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

OKBM Afrikantov

Country of origin

Russian Federation

Reactor type

PWR

NPS electrical capacity in cogeneration mode (MW(e))

6

NPS thermal capacity in cogeneration mode (MW(th))

14

NPS electrical capacity in condensation mode (MW(e))

9RP thermal capacity (MW(th))

38

Expected capacity factor (%)> 70*

Design life (years)

40

2Plant footprint (m )20000*

Coolant/moderator

Light water

Primary circulation

Natural circulation

System pressure (MPa)16.2oCore inlet/exit temperatures ( C)

250 /325

Main reactivity control mechanism

Control Rod Driving Mechanism (CRDM)

RPV height (m)6

RPV diameter (m)2.4

RPV weight (metric ton)

88.2

Configuration of reactor coolant system

Integral type

Power conversion process

Indirect Rankine cycle

Fuel type/assembly array

UO2/hexagonal

Fuel assembly active length (m)0.9Number of fuel assemblies

121

Maximum fuel enrichment (%)

19.7

Fuel burnup (GWd/ton)

N/A

Fuel cycle (months)

120-144

Cogeneration capability

Yes

Approach to engineered safety systems

Passive

Number of safety trains

2

Refuelling outage (days)

Refueling is included in the integrated maintenanceDistinguishing features

Natural circulation in the primary circuit

Modules per plant

1*

Target construction duration (months)

54 - 60*

Seismic design (Richter scale points)

7*

Core damage frequency (per reactor-year)

10E–6

Design Status

Final design* - The values are given for the basic design. If the design is adjusted for a specific site, theshown values may change.
Reactor Pressure Vessel and Internals
The RPV is a welded cylindrical “container” with an elliptical bottom. At the top of thevessel there are pipes for feedwater supply and superheated steam removal, as well as thosefor the connection of the primary circuit systems and the auxiliary process systems. The RPV
197

head consists of a load-bearing slab, a shell attached to this slab and sealed by a weld, and a
top slab welded to the shell. The cavity between the top slab and the load-bearing slab isfilled 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 loadbearing slab and penetrate through the cover. Points of penetration through the top slab are
sealed. Fuel assemblies are located in 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 fuel assemblies into CPS
control rods and connecting the CPS control rods to CPS drives.
Reactor Coolant System
Core heat removal is based on conventional two-circuit methodology. The core is cooled andmoderated by water through natural circulation of coolant in the primary circuit. Hot coolant
is cooled in a once-through steam generator, where slightly superheated steam is generated,then supplied to the turbine. This design eliminates large-diameter pipelines in the primarycircuit 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 vertical surface-type heatexchanger 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 each section is carried out through the pipes in the reactor vessel. Counter flow
circulation is used, i.e., the primary circuit coolant moves downward in the inter-tube space,
while the secondary circuit coolant is moved upward in the tubes. In case of inter-circuit
leaks, it is possible to cut off any section automatically or remotely. Identification of theleaking section is carried out with the use of the detection blocks of the radiation and processcontrol 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 publicarea and at the same time far-off location from main technical bases, which could provide
timely technical support. In view of its small power the emergency systems are simpler andoften do not require active systems performance. Land-based and floating power units use theadvanced active and passive safety systems for emergency cooling over an unlimited time
during design-basis and beyond design-basis accidents. Low thermal capacity of reactor
allows using natural circulation at primary coolant circuit and passive safety systems as
primary safety systems. The auto-protective 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 emergency cooldown system (ECDS) is used toremove decay heat. This system functions on natural physical processes and - because there isan air heat exchanger cooled by the atmospheric air - ensures that the decay heat is being
removed from the reactor for an unlimited time in all types of accidents. Because of this, andconsidering the measures taken to enhance the reliability of the passive ECDS, there are no
active ECDS channels in the ABV 6E reactor design, which allows the output of emergency
198

power supply sources to be reduced. The passive ECDS is made of two independent channels
connected to two SGs each. Either channel, independently of the operability of the otherchannel, is capable of performing the ECDS 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 primarycoolant leak and to cool the reactor core in case of LOCA. The ECCS comprises of the highhead pumps that inject water into the RPV if power supply is available, and the hydroaccumulators that supply water under the action of the compressed gas.
Containment Systems
The metal-and-water shielding (MWS) tank is a substantial structure for the equipment of the
RP. RPV, two pressurizers and the cooler of the purification and heat removal system areenclosed inside the dry caissons of the MWS tank. The passive reactor caisson water flooding
system is designed to protect the RPV against melt-down in severe beyond-design-basis
accidents associated with core damage. The system feeds the primary coolant condensate to
the RPV caisson. It is also possible to supply water from the fresh water intake and pumping
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 26000 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 withboth 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 towithstand the 10-ton aircraft crash. As the analysis of emergencies has shown, the radiationand 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. This impact is also limited in beyond-design-basis accidents.
7. Design and Licensing Status
The final design of ABV-6E has been accomplished. The design has not been licensed yet.
8. Plant Economics
The average specific cost of construction for the serial power unit with ABV-6E is around8,500 $/kW. The contribution of the serial power unit to the prime cost of electricity is around0.14 $/kW·h, The cost characteristics are shown without taking into account the costs of thecoastal structures and plant infrastructure, because they depend, in a great extent, on the site
(geo-physical data, climatic conditions, applicable local regulatory acts, social-economic andinfrastructure conditions, features of the selected operation model and other factors).
9. Plant Arrangement

199

200

Figure 4: General plan of the FNPP (Reproduced courtesy of OKBM Afrikantov)

201

Figure 5: General plan of the land-based NPP (Reproduced courtesy of OKBM Afrikantov)

SHELF (NIKIET, Russian Federation)
1. Introduction
A power unit based on the SHELF reactor is designed for use as a local power source forusers in remote and hard-to-reach locations. The power unit, as shown in Figure 1 is a power
capsule to generate 6.4 MW(e). The power capsule accommodates all reactor components,
the turbine generator package (TGP), the automated and remote control system, monitoringand protection system and the electricity output regulation. SHELF power capsule can beused for both floating and submerged nuclear power plants. Engineering design of SHELF issimilar to a large degree with marine propulsion nuclear power plants. The power unit isdelivered as a single item with all of its components accommodated inside a high-strength
vessel (containment). This ensures high quality of the module fabrication at a specializedmachine-building plant. The reactor is refueled, exhausted equipment is replaced, and thereactor, the TGP and other components are dismantled at the end of service life at a
specialized plant. The capability for long-term unattended automated operation of the reactor,
the TGP and other equipment inside the SHELF module eliminate the need of plant operators
inside the power unit and to keep the module unmanned during the automated operation
period equal to 5000 hours. After the automated operation, the scheduled maintenance isconducted within 15 days.

Figure 1: Power unit of an NPP with SHELF-10 reactor facility(Reproduced courtesy of NIKIET)
2. Target Application
SHELF reactor is designed for use as a source of power for installations in remote and hardto-reach locations with decentralized power supplies including the sites on Arctic shore
regions. The electric power of the single-unit plant with SHELF reactor is 6.4 MW while202

thermal power is 28 MW. Depending on the consumer requirements a SHELF plant can beequipped with a system for direct heat supply to resident and production premises with acapacity of 12 Gcal/h or a distillation desalination plant with a capacity of 500 m3/h of fresh
water. In case of land based plant, the heat is rejected by external heat exchangers withmechanical pumping of air. The SHELF plant does not require local water sources. Decay
heat removal in case of submerged deployment is provided by sea water cooling.
3. Development Milestones
2012
2015
2016

A concept of an undersea power unit with the SHELF reactor developed by NIKIET. The
request for proposal was issued
Study of a land-based unit design and the development of a line of integral facilities in anelectric power range from 1 to 8 MW.
Detailed design of SHELF reactor unit with an electric power of 5.4 MW underway at
NIKIET4. General Design Description
SHELF is a water-cooled reactor of integral layout with a combined (forced and natural)
coolant circulation modes. The fuel cycle is 6 years with the design capacity factor of 80%.
The components for SHELF are accommodated inside a cylindrical power capsule with an
inner diameter of 8 m and a length of 14 m. The reactor equipment is installed in the rear
portion of the capsule and the TGP equipment is installed in its front. Besides, there is acompact module housing the auxiliary systems including an air conditioning system formodule room temperature maintenance at a level of not more than 50°C. The unit’s external
systems include an automated instrumentation and control system (AICS), uninterrupted
power supply systems, as well as the reactor facility and TGP auxiliary systems, including aventilation system, a negative pressure system and others. Following the delivery to theoperation site, the SHELF unit module is installed, secured on dedicated supports andconnected to the other NPPs equipment through service lines. The modular approach allows
the capital construction cost to be minimized and the requirements to the nuclear plant
deployment site reduced.
Reactor Core
The core is of the heterogeneous cartridge type and consists of 163 hexahedral fuelassemblies (FA) of three different types. In a number of FAs burnable poison and control
absorber rods are located. Absorber rods for reactivity control are united in six identical shimgroups. The choice of fuel composition was governed by the maximum fuel load needed toprovide the required reactor core life with the fuel enriched to not more than 20%; fuelcomposition consists of uranium dioxide in a silumin matrix in the form of cylindrical fuelelements. The specific advantages offered by this fuel type in comparison with the ceramicfuel of the container type applied in large NPP are as follows:• A potentially higher nuclear fuel burn-up (up to 160 MW∙day/kg);• Low temperature in the fuel composition and, as a result, a small amount ofaccumulated heat;• Serviceability ensured at different power levels in a combination with highmanoeuvrability;• An extra safety barrier due to the fuel matrix properties, ensured relative to thefission product release.
The reactor core size, composition and load are chosen so to provide the preset core life andcapability conditions. Highly pure water is used as the core coolant and moderator. The
ammonia addition is used to generate hydrogen in water with the aim to prevent corrosive203oxidative radiolysis products generation. The operating time of the reactor with one fuel load
is 5.6 years, with no scheduled outages for maintenance included, and the reactor core life is
40000 hours. The reactor core exhibits negative temperature, density and power reactivity
coefficients across the range of the reactor parameters variation. This factor is favourable forthe core self-regulation and the safety improvement. Reactor refueling according to theadopted concept is performed on the specialized enterprise basis. In-service fuel handling is
not envisaged.
Reactor coolant system
SHELF’s thermal-hydraulic circuit consists of two self-sustained systems: nuclear steam
supply system (NSSS) and a turbine generator system. NSSS comprises the following major
systems:
•
•
•
•
•
•
•
•
•
•

Primary circuit system;
Secondary circuit system;
Emergency core cooling system (ECCS);
Emergency cooldown system (ECS);
Makeup, dual and emergency absorber injection system;
Equipment cooling system;
Reactor overpressure protection system;
SG overpressure protection system;
Safeguard vessel and containment overpressure protection system;
Instrumentation and control system (I&C).

SHELF is an integral-type reactor. This ensures the compactness of the system and equipment
and improves the reliability of the facility due to a smaller number of supply lines subjectedto the primary coolant pressure. The simplicity of the coolant line layout, combined with the
integral reactor design, provides, when the coolant circulates naturally, for a high level ofremoved power being enough to keep the reactor operating at a power of up to 65% of the
rated value in the event the pump power supply loss and for the accident-free core cooldown.

Figure 2: Key systems of the SHELF reactor facility (Reproduced courtesy of NIKIET)

204

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer
Country of origin

NIKIET, Research and Development Institute of Poweri
i
RussianFederation

Reactor type

Integral PWR

Electrical capacity (MW(e))

6.4

Thermal capacity (MW(th))

28

Expected capacity factor (%)

80

Continuous operation period (hr)

40000

Design life (years)
Plant footprint (m2)

N/A

Coolant/moderator

Light water

Primary circulation

Forced and natural circulation

System pressure(MPa)–oCore inlet/exit temperatures ( C)–Core flow rate (ton/hr)–

Steam generating capacity (ton/hr)–o

Steam temperature ( C)–

Pressure (MPa abs)–
2

Neutron fluence at the end of life time (neutron/cm )–Main reactivity control mechanism

Control Rod Drive Mechanism (CRDM)

RPV height (m)14

RPV diameter (m)8Module weight (metric ton)–

Configuration of reactor coolant system

Integral PWRPower conversion process–

Fuel type/assembly array–

Fuel assembly active length (m)–Number of fuel assemblies

163

Fuel enrichment (%)< 20

Fuel burnup (GWd/ton)–

Fuel cycle (years)

5.6

Cogeneration capability

Yes

Approach to engineered safety systems

Combined active and passive

Number of safety trains–

Refuelling outage (days)–

Distinguishing features

Power source for users in remote and hard-to-reach locations;can be used for both floating and submerged nuclear power
plantsModules per plant–

Target construction duration (months)–

Seismic design (g)–Core damage frequency (per reactor-year)–

Design Status

Detailed design underway

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Pressurizer
The design calculations have shown that the primary circuit pressure does not exceed thespecified level in the design-basis situations. The design includes an overpressure protection
system as an additional engineering approach for the management of beyond design-basis
accidents and serves to prevent, in such cases, the primary circuit system boundaries frombreaking down or being loaded with forces in excess of those permitted. The system
incorporates two parallel initiation lines. To confine radioactive products, the fluid isdischarged over the reactor overpressure protection lines, via a bubbler, into the biologicalshielding inner tank inside the safeguard vessel. Apart from the primary circuit pressurizing
system, the SHELF facility with an integral reactor contains independent pressurization andsecondary circuit equipment overpressure protection systems, and the steam generator,
safeguard vessel and containment overpressure protection systems.
Power conversion system
The SHELF two-circuit integral water-cooled reactor facility is an integration of systems andcomponents intended to convert nuclear energy to thermal power. It includes a reactor andassociated systems required for its normal operation, emergency cooling, emergency
protection and maintenance in a safe condition. Auxiliary and support functions are
performed by other SMR plants. The purpose of the primary circuit system is to remove heatfrom the reactor core and transfer it to the secondary circuit fluid in the SG. It comprises a
main circulation line and a coolant inventory variation compensation system. The purpose of
the secondary circuit system is to generate superheated steam from feedwater and transfer
heat to the NPP’s turbine generator plant. The reactor facility’s secondary circuit systemcomprises a steam generator installed inside the reactor vessel and pipelines with valves(outside the vessel).
Control and Protection System
A monitoring, control and protection system (CPS) is used to perform the functions of thereactor facility’s control and protection in all normal and emergency operation modes. The
CPS is a part of the SHELF unit’s automated instrumentation and control system (AICS). The
CPS is designed to:
•
•
•
•
•

Control the reactor core reactivity and power through the generation andperformance of control actions on the reactor core in all operating modes of thereactor facility;
To monitor the neutron flux (power), its variation rate, and the process parameters
required for the protection and control of the reactor core reactivity and the reactor
facility power;
To function as the initiating portion (at the upper coordinating level) of the control
safety systems and to regulate the reactor facility’s key process parameters;
To render and keep the reactor subcritical.
The requirements to the CPS are defined both by Russian and international
regulatory documents.

Steam Generator
The once-through steam generator (SG) is part of the SHELF reactor and is designed togenerate superheated steam in the process of the reactor operation and to remove heat fromthe primary circuit system during the reactor cooldown. The SG comprises a tubing,collection and distribution chambers installed in the annulus above and below the tubes,
steam and feedwater lines installed inside the reactor vessel, steam and water risers installedon the reactor vessel flange, and outside steam and feedwater lines with shutoff and isolation

206

valves. The SG’s heat transfer surface is divided into four independent sections cut off, whenrequired, by the shutoff and isolation valves.
5. Safety Features
High level of the SHELF reactor safety is achieved through the following aspects.
1. Use of an integral water-cooled water-moderated reactor with well-developed intrinsic
self-protection properties and the following inherent features:•••

negative core reactivity coefficients across the operating range of parameters;
high level of natural coolant circulation, which ensures efficient cooling of andheat transfer from the core in beyond design-basis and design-basis accidents;
high heat accumulating capacity of metal structures and large coolant inventory in
the reactor, which ensure relatively slow transients during accidents caused by theloss of normal heat removal from the core.

2. A defense-in-depth system of barriers to the spreading of ionizing radiation andradioactive products of uranium fission into the environment, as well as a combination of
implemented engineering and organizational measures to protect these barriers frominternal and external impacts. Safety barrier system includes:
•
•
•
•
•
•

Fuel matrix;
Fuel cladding;
Leak tight primary circuit – reactor vessel;
Safeguard vessel;
Confinement valves;
Containment.

3. Application of passive safety systems and features that operate based on the naturalprocesses without energy supply from outside. Such systems are as follows:
•
•
•
•
•
•

Structure of CPS drive actuators;
Decay heat removal system (DHRS);
Emergency core cooling system (ECCS);
Safeguard vessel enables the reactor core to maintain primary coolant level and to
remove heat under all serious accidents as well as the radioactive productconfinement when the primary circuit leak tightness is lost;
Containment that restricts the radioactive product release when the safeguard
vessel is opened for some activities and in case of beyond design-basis accidents;
Reactor, safeguard vessel and containment protection systems equipped withmembrane and rupture devices that operate under direct fluid action when apressure exceeds the allowed limit.

4. Safety system reliability:
High reliability level of safety systems is achieved through implementing the following
principles:••

Passive operation not requiring any actions for initiation;
A diversity of safety systems and devices achieved through the use of different
operating principles of systems (e.g., use of the CPS electromechanical drives forthe emergency shutdown of the reactor, these being functionally divided into
reactivity compensation member drives and emergency protection drives, as well
as of the liquid absorber injection system);

207

•
•

Redundancy of safety systems;
Continuous and periodic monitoring of the in-service status of systems andcomponents.

5. Protection against external impacts:
The reactor facility’s containment ensures that the reactor components inside the containment
and the safeguarded vessel are not damaged in the event of external impacts, including
typhoon, hurricane, snow and icing, as well as of helicopter or airplane impacts on the
SHELF NPP.
6. Plant Operational Performance
The electric power of one SHELF unit is 6.4 MW(e), and the thermal power is 28 MW(th).
The current supplied to the consumer system is alternate and three-phase (voltage 0.4 kV ±2%, frequency 50 Hz ± 1 Hz). The nuclear plant base operation mode is power operation in arange from 20 to 100% full power with the capability to vary the consumed power daily andannually. The power increase and decrease rate is 1% (forced primary coolant circulation)and 0.3% (natural primary coolant circulation). The time of the reactor operation with one
fuel load is 40000 effective hours.
7. Instrumentation and Control System
The automated process control system (APCS) of a nuclear plant with the SHELF-10 reactor
is intended to control the major and auxiliary electricity generation processes in all modes of
the unit operation:
1. Normal operation:
•
•
•
•
•
•

Phased automated initiation;
Operation at steady power levels in a range of 20 to 100% Nnom with forced primarycoolant circulation;
Operation at steady power levels in a range of 20 to 95% Nnom with natural primarycoolant circulation;
Switchovers from one steady power level to another in the above power ranges at apreset rate (manoeuvring);
Switchover from natural primary coolant circulation to forced circulation;
Scheduled automated deactivation.

2. Anticipated operational occurrences:••

Emergency power reduction and operation with a decreased steam supply due tofailures of the reactor facility’s key components or feedwater supply and steam receipt
systems;
Deviation of the reactor facility’s parameters to beyond the preset limits, leading to
the need for the operator intervention or automatic controlling actions to normalize
the parameters.

3. Emergency: Emergency deactivation in the event of reactor facility parameters deviating
beyond the safe operation limits or in the event of equipment failures leading to the safe
operation limits being violated.
8. Plant Arrangement
The undersea power unit module is an energy capsule which accommodates all componentsof the reactor facility, the TGP, and the unit equipment automated and remote control,

208

monitoring and protection systems, including the electricity output regulation, monitoring andcontrol equipment. The unit’s land based installation includes the AICS equipment, the AICS
undersea equipment uninterruptible power supply systems, as well as auxiliary systems to
support the reactor facility and TGP operation, including the ventilation system, the negative
pressure system and others. The power supply system arrangement is selected based on the
deployment conditions and the consumed auxiliary power with regard for the supporting
utilities and physical protection systems. The power generation system consists of a generator
connected to a steam turbine, generator current distributors and switchgear. The plant,
including its power generation and supply system, and the auxiliary power system, isoperated automatically with control from the APCS.
9. Design and Licensing Status
The licensing of the nuclear plant design is scheduled for 2017-2019.
10. Plant Economics
The current estimated performance indicators prove the nuclear plant with the SHELF reactor
facility to be highly competitive not only in comparison with similar nuclear plant designs but
also as compared to conventional generation based on imported fuel. The estimated payback
period for a single-unit nuclear plant with the SHELF reactor facility is 14 years.

209

210

HIGH TEMPERATUREGAS COOLEDSMALL MODULAR REACTORS212

HTR-PM (Tsinghua University, China)
1. Introduction
In 1992, the China Central Government approved the construction of the 10 MW pebble bed
high temperature 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 completed a lot of experiments on the HTR-10 to verify
crucial inherent safety features of modular HTRs, including:••••

Loss of off-site power without scram;
Main helium blower shutdown without scram;Withdrawal of control rod without scram;
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 (HTR-PM) project was launched. The first concrete of
the HTR-PM demonstration power plant was poured on December 9, 2012, in Rongcheng,
Shandong Province, as shown in Figure 1(a). According to its planned schedule, the power
plant will start power generation in 2017 [1]. The civil work of the nuclear island’s buildings
has been finished as shown in Figure 1(b), and the first of two reactor pressure vessels was
installed in March 2016.

Figure 1: Construction of HTR-PM on (a) December 9, 2012 and (b) May 25, 2015
(Reproduced courtesy of INET)
2. Target Applications
The HTR-PM is a commercial demonstration unit for electricity production. The twin reactor
modules driving a single turbine configuration was specifically selected to demonstrate its
feasibility.
Following HTR-PM demonstration plant, commercial deployment of HTR-PM based onbatch construction is foreseeing, and units with more modules and bigger power size areunder design. Units with multiple standardized reactor modules coupling to one single steam
turbine (such as 200, 600 or 1000MW) are envisaged.
A conceptual design has been finished for a 600 MW(e) multi-module HTR-PM nuclear
power plant, which consists of six reactor modules coupling to one steam turbine. Each213

reactor module has the same design as the HTR-PM demonstration plant, with independent
safety systems and shared non-safety auxiliary systems. The footprint of a multi-module
HTR-PM plant is not remarkably different from that of a PWR plant generating the same
power. Figure 2 shows a 2 × 600 MW(e) HTR-PM nuclear power plant for cogeneration.
Research and development on process heat applications, hydrogen production and gas
turbines are continuing for future application

Figure 2: The 2 × 600 MW(e) HTR-PM multi-modules plant (Reproduced courtesy of INET)
3.

Development Milestones
2001

Launch of commercial HTR-PM project

2004

Standard design of HTR-PM started

2006

HTR-PM demonstration power plant approved as one of National Science and Technology Major Projects

2006

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. and Tsinghua University

214

2006-2008

Basic design of HTR-PM completed

2009

Assessment of HTR-PM PSAR completed

2012

First Pour of Concrete of HTR-PM

2013

Fuel plant construction started

2014

Qualification irradiation tests of fuel elements completed

2015

Civil work of reactor building finished2016

RPV and core barrel etc. delivered, installation of main components ongoing

2017

First operation expected

4. General Design Description
The HTR-PM should achieve the following technical goals: (1) demonstration of inherent
safety features; (2) demonstration of economic competitiveness; (3) confirmation of proventechnologies; (4) standardization and modularization.
Design Philosophy
The HTR-PM consists of two pebble-bed reactor modules coupled with a 210 MW(e) steam
turbine [2], as shown in Figure 3. Each reactor module includes a reactor pressure vessel(RPV); graphite, carbon, and metallic reactor internals; a steam generator; and a main helium
blower. The thermal power of each reactor module is 250 MW(th), the helium temperaturesat the reactor core inlet/ outlet are 250/750 °C, and steam at 13.25 MPa/567 °C is produced at
the steam generator outlet.

Figure 3: The HTR-PM demonstration nuclear power plant: (a) Front view; (b) top view
(Reproduced courtesy of INET)
Reactor Core and Power Conversion Unit
The primary helium coolant works at the pressure of 7.0MPa. The rated mass flow rate is
96kg/s. 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 level where it reverses the flow direction and flow into the pebble 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 isheated up in the active reactor core and then is mixed to the average outlet temperature of
750ºC and then flows to the steam generator.
As shown in Figure 4, the reactor core and the steam generator are housed in two steelpressure vessels which are connected by a connecting vessel. Inside the connecting vessel,
the hot gas duct is designed. All the pressure retaining components, which comprise theprimary pressure boundary, are in touch with the cold helium of the reactor inlet temperature.

215

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer:

INET, Tsinghua University

Country of Origin:

China

Reactor Type:

Modular Pebble bed High Temperature Gas-cooled Reactor

Electrical Capacity (MW(e)):

210

Thermal Capacity (MW(th)):2x250

Expected Capacity Factor (%):

85

Design Life (years):

40

2

Plant Footprint (m ):~60000 (site)

Coolant/Moderator:

Helium / Graphite

Primary Circulation:

Forced circulation

System Pressure (MPa):

7

Main Reactivity Control Mechanism:

Negative temperature coefficient; control rod insertion

RPV Height (m):25

RPV Diameter (m):5.7 (inner)

Coolant Temperature, Core Outlet (°C):

750

Coolant Temperature, Core Inlet (°C):

250Power Conversion Process:Indirect Rankine Cycle

High-Temp Process Heat:

Yes, possible with different configuration

Low-Temp Process Heat:

Yes, possible with different configuration

Cogeneration Capability:

Electricity only; possible with different configuration

Design Configured for Process Heat Applications:

Active Safety Features:No
Yes, inherent safety by large negative temperature coefficients,
large heat capacity
Yes, control rod insertion with SCRAM; Turbine trip

Fuel Type/Assembly Array:

Spherical elements with coated particle fuel

Fuel Pebble Diameter (cm):6

Number of Fuel Spheres:

420000 (in each reactor module)

Passive Safety Features:

Fuel Enrichment (%):

8.5

Fuel Burnup (GWd/ton):

90 (in average while discharge)
N/A; Online / on-power refuelling
Fuel remain in the reactor for ~35 months
Control rod insertion; Circulator trip; isolation of secondary
circuit; drain of steam generator in the case of steam generator

Fuel Cycle (months):
Emergency Safety Systems:
Residual Heat Removal Systems:

Passive

Refuelling Outage (days):

N/A since online

Distinguishing Features:

Inherent safety, no need for offsite emergency measures

Modules per Plant:

Two modules with their own steam-generator feeding one
turbine-generator set.

Estimated Construction Schedule (months):

59 months for the demonstration power plant

Seismic design (g):0.2

Predicted large release frequency:

Core damage frequency not applicable to HTGRs
No off-site shelter or evacuation plan needed

Design Status:

Under construction.

216

Figure 4: HTR-PM primary loop (Reproduced courtesy of INET)
Fuel characteristics

Figure 5: Spherical Fuel Elements(Reproduced courtesy of INET)

Fuel elements are spherical ones. Every fuelelement contains 7g heavy metal. The
enrichment of U-235 is 8.5%. Uranium
kernels of about 0.5mm diameter are coatedby three layers of pyro-carbon and one layerof silicon carbon. Coated fuel particles are
dispersed in matrix graphite which is 5cm indiameter. Surrounding the fuel containing
graphite matrix is a 5mm thick graphite layer

Fuel handling system
The operation mode of HTR-PM adopts continuous fuel loading and discharging: the fuelelements drop into the reactor core from the central fuel loading tube and are discharged
through a fuel extraction pipe at the core bottom. Subsequently, the discharged fuel elements
pass the burn-up measurement facility one by one. Depending on their state of burn-up they217

will either be discharged and transported into the spent fuel storage tank while running up to
the design threshold of burn-up, or be 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 graphitic side reflector. Reactivity control isachieved through 24 control rod assemblies, and six SAS shutdown systems serve as areserve shutdown system. The control rods are used as a regulating group that is used during
normal plant operation to control reactivity, but also for hot shutdown. The SAS system isused for long-term shutdown. 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 is all
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 inner graphite reflector andouter 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 protected against high temperatures from the core by the cold helium borings of the side
reflector, which act like a shielding temperature screen.
5.

Plant Safety Features

Description of safety concept
The HTR-PM will realize the following safety features:(1) The radioactive inventory in the primary helium coolant is very small when the reactors
are working at normal operation conditions. Even if this limited amount of radioactivity
would be released into the environment following an accident, there is no need to take any
emergency measures. (2) For any reactivity accident or for a loss of coolant accident, the riseof the fuel element temperature will not cause a significant additional release of radioactive
substances from the fuel elements. (3) The consequences of water or air ingress accidents
depend on the quantity of such ingresses. The ingress processes and the associated chemical
reactions are slow, and can readily be terminated within several dozens of hours (or evensome days) by taking very simple actions.
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 ensuresthat the fundamental safety functions are maintained. A large negative temperature
coefficient, large temperature margin, low excess reactivity (due to on-line refuelling) andcontrol rods ensures safe operation and limits 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 fueltemperature 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 [3].
Another feature of the HTR-PM design is the long time period of accident progression due tothe large heat capacity of fuel elements and graphite internal structures. It requires days forthe fuel elements to reach the maximum temperature when the coolant is completely lost.

218

For the HTR-PM plant, the most significant design basis accident is the complete loss ofprimary helium coolant. In such an accident scenario, no any additional cooling systems aredesigned for cooling the reactor core. The decay heat in fuel elements shall dissipate to theoutside of the pressure vessel primarily by means of heat conduction and radiation within thereactor core internals. The maximum fuel temperature during this accident scenario is thehighest and most challenging for the fuel, therefore this accident scenario is the most criticalone for the HTR-PM design. In the case of HTR-PM, the maximum fuel temperature will beless than 1620°C for any conceivable accident.
Engineered Safety System Configuration and Approach
When accidents occur, a very limited number of reactor protection actions shall be called
upon by the reactor protection system. No or very limited actions through any systems orhuman interventions are foreseen after the limited reactor protection actions are activated.
The limited reactor protection actions shall be to trip the reactor and the helium circulator, toisolate the primary and secondary systems. When there is large leak or rupture of steam
generator heat exchange tubes, a water discharge system is designed to minimize the amountof water ingress into the reactor core.
Reactor Cooling Philosophy
Normally the reactor is cooled by steam generating system.
Under accident conditions, the primary helium circulator shall be stopped automatically.
Because of the low power density and the large heat capacity of the graphite structures, thedecay 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 limits. The decay heat shall be removed to heat sink passively
by reactor cavity cooling system (RCCS). Even if the RCCS fails, the decay heat can beremoved transferring through the concrete structure of reactor cavity while the temperaturesof fuel elements are under design limit.
Containment Function
Retention of radioactivity materials is achieved through multi-barriers. The fuel elementswith coated particles serve as the first barrier. The fuel elements used for HTR-PM have been
demonstrated to be capable of retaining fission products within the coated particles undertemperatures of 1,620ºC which is not expected for any plausible accident scenarios. The
second barrier is the primary pressure boundary which consists of the pressure 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 building
and some additional auxiliary buildings which house primary helium containing components.
6. Plant Operational Performances
The HTR-PM demonstration power plant is under construction. Due to online fuel loading
mode of HTR-PM, better availability factor can be expected compared with other power
plants operating in a mode 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.

219

8. Plant Arrangement
The nuclear island contains reactor building, nuclear auxiliary building, spent fuel storage
building and I&C building, as shown in Figure 6. The steam turbo-generator, which is similar
to that of a conventional fossil-fired power plant, is housed in the turbine building.

Figure 6: Nuclear island and conventional island layout (Reproduced courtesy of INET)
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.
10. Plant Economics
The HTR-PM demonstration power plant is a first-of-a-kind project in the world. Based on
the cost data of the demonstration plant, a preliminary estimation shows that the cost offollowing six-module HTR-PM power plant is a little higher than that of a PWR plant
generating the same power. However, the difference is only around 10~20% [4].
[1]

[2]
[3]
[4]

220

Zhang, ZY; Dong, YJ; Li, F; Zhang, ZM; Wang, HT; Huang, XJ; Li, H; Liu, B; Wu,
XX; Wang, H; Diao, XZ; Zhang, HQ; Wang, JH. The Shandong Shidao Bay 200
MW(e) High-Temperature Gas-Cooled Reactor Pebble-Bed Module (HTR-PM)
Demonstration Power Plant: An Engineering and Technological Innovation.
Engineering 2 (2016) 2(1): 112–118.
Zhang, ZY; Wu, ZX; Wang, DZ; Xu, YH; Sun, YL; Li, F; Dong, YJ. Current status
and technical description of Chinese 2 x 250 MW(th) HTR-PM demonstration plant.
NUCLEAR ENGINEERING AND DESIGN, (2009) 239(7) : 1212-1219
https://www.technologyreview.com/s/600757/china-could-have-a-meltdown-proofnuclear-reactor-next-year/
Zhang, ZY; Sun, YL. Economic potential of modular reactor nuclear power plants
based on the Chinese HTR-PM project. NUCLEAR ENGINEERING AND DESIGN,(2007)237: 2265–2274

GTHTR300 (Japan Atomic Energy Agency,
Japan)
1. Introduction
GTHTR300 (Gas turbine High Temperature Reactor 300 MW(e)) is a multi-purpose,
inherently-safe and site-flexible SMR (small modular reactor) that Japan Atomic Energy
Agency (JAEA) is developing for commercialization in around 2020s [1]. As a GenerationIV technology, the GTHTR300 offers important advances comparing to current light water
reactors. The reactor coolant temperature is significantly higher in the range of 850-950oC.
Such high temperature capability as proven in JAEA’s HTTR test reactor operation enables awider range of applications such as high temperature heat applications. The design employs adirect-cycle helium gas turbine to simplify the plant by eliminating water and steam systems
while delivering 45-50% generating efficiency comparing to about 33% efficiency by current
reactors [2]. The design incorporates all ceramic fuel, low power density but high thermal
conductivity graphite core, and inert helium coolant to secure inherent reactor safety. The
inherent safety permits siting proximity to customers, in particular to industrial heat users so
as to minimize the cost and loss of high temperature heat supply. Dry cooling becomes
economically feasible due to the use of gas turbine. The waste heat from the gas turbine cycle
is rejected from 200oC, creating a large temperature difference from ambient air making thedry cooling tower size per unit of power generation comparable to the wet cooling towers
used in nuclear plants today. The economical dry cooling permits inland and remote reactor
siting even without a large source of cooling water.

GTHTR300

Reactor building

Figure 1: Reactor System Configuration of GTHTR300 (Reproduced courtesy of JAEA)
2. Target Applications
Typical applications include electric power generation, thermochemical hydrogen production[3], desalination cogeneration using waste heat only, and steelmaking. The reactor thermal
power may be rated up to 600 MW(th) maximum. The maximum product output per reactor
is 120 t/d hydrogen enough to fuel about one million cars, 280-300 MW(e) electricity
generation with additional seawater desalination cogeneration of 55000 m3/d potable water
for about a quarter million of population, and annual production of 0.65 million tons of steel.
All these are produced without CO2 emission.

221

Reactor

Hydrogen production

Reactor : 600 MWt
Hydrogen : 120 t/d
Figure 2: Cogeneration plant layout for hydrogen (left) and for desalination (right)(Reproduced courtesy of JAEA)
3. Development Milestones
2003

Basic design

2004

Design development start

2005

Cogeneration design (GTHTR300C)2014

IS process continuous H2 production test facility construction

2015

Basic design for HTTR-connected gas turbine and hydrogen test plant (HTTR-GT/H2)

2020-25

HTTR-GT/H2 test operation (planned)

2025

Construction of lead commercial plant (planned)

4. General Design Description
Design Philosophy
The overall goal of the commercial plant design is to provide a family of system options
capable of producing competitive electricity, hydrogen, desalination, other products, and yet
deployable in the near term. The development of the multiple systems simultaneously does
not necessarily suggest having investment and risk multiplied. Rather, the development
requirement is minimized thanks to a design philosophy of system simplicity, economic
competitiveness and originality, namely the SECO philosophy. There are three major
elements to this design philosophy.
The first element is technology simplification. All design variants are built on the premisethat they share common system technologies to the maximum extent possible. As a result, thedesign variants share a unified reactor and primary coolant circuit, an aerodynamically andmechanically similar line of helium gas turbines used for electricity production, and the IS
process selected to produce hydrogen [4]. The helium gas turbine and the IS process arecompatible application systems with the high temperature reactor heat source to enableeconomically competitive energy production.
The second element of the SECO design philosophy has been incorporating unique design
attributes that are less demanding on the system technologies required. The efforts in this area
have resulted in such original design simplification as conventional steel reactor pressure
vessel construction, horizontal gas turbine installation, and system modular arrangement
among others as shown in Figure 3. The technical design features of the GTHTR300 plantvariants in greater detail.

222

The third element that has been made possible by constant pursuit of technology and design
simplification is a focused technological development scope that comes with low risk andinvestment of overall development. Furthermore, since the technologies to be developed areshared by several systems, the benefit of investing in any one development is increased. On
the site where the HTTR is constructed for acquiring the reactor technology, JAEA has also
been carrying out research and development on the helium gas turbine and the IS process.
Design Features
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 threefunctionally-oriented pressure vessel units, housing the reactor core, the gas turbine, and theheat exchangers respectively. The multi-vessel system facilitates modular construction andindependent 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 in a GTHTR300C design that can accept various roles ofcogeneration while sharing equipment designs with GTHTR300.
ReactorElectric generator

Gas turbine
Recuperator& Precooler

Figure 3: Functionally-oriented pressure vessel units of the reactor primary system(Reproduced courtesy of JAEA)
Fuel Design
The fuel design is coated fuel particle of less than 1 mm in diameter as shown in Figure 4.
Each particle consists of a UO2 kernel coated by four layers of low and high density pyrocarbon and silicon carbide. The all ceramic particle fuel is heat resistant up to 1600oC.
Approximately ten-thousand 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 of a hexagonal graphite fuel block of about 1 m long and 41 cm across, where
the annulus formed between the fuel 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.

223

Fuel Block

Fuel Rod

Fuel Compact(UO2)2-YSZ）
Kernel（PuO
Buffer Layer
iPyC Layer
SiC Layer

Coated Fuel Particle

oPyC Layer

Figure 4: Fuel design features (Reproduced courtesy of JAEA)
Design Development
While the reactor technologies required for the GTHTR300 are developed mainly withconstruction and operation of JAEA’s 30 MW(th) and 950oC test reactor, we are separately
developing and testing all key balance of plant technologies needed for the commercial SMR,
including test validation of the helium gas turbine equipment at one-third to full scale,production-scale fuel fabrication lines, thermochemical hydrogen production process, and thesuperalloy heat exchanger capable of transferring 950oC reactor heat to the hydrogenproduction process.

Figure 5: Plant technologies developed for GTHTR300 (Reproduced courtesy of JAEA)

224

AJ
MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology developer

JAEA jointly with MHI, Toshiba/IHI, Fuji Electric, KHI, NFI, etc.

Country of origin

Japan

Reactor type

Prismatic HTGRElectrical capacity (MW(e))

100~300 MW(e)Thermal capacity (MW(th))< 600 MW(th)Design capacity factor>90%

Design life (years)

60

Coolant

heliumModerator

graphite

Primary circulation

forced circulation

System pressure

7 MPa

Reactivity control mechanism

control rod

RPV height (m)23

RPV diameter (m)8Coolant temperature, core Outlet (°C)

850-950

Coolant temperature, core Inlet (°C)

587-633

Integral design

No

Power conversion process

direct Brayton cycle

High temperature process heat

Yes

Low temperature process heat

Yes

Cogeneration capability

Yes

Design configured for process heat applications

Yes

Safety features

Inherent

Fuel type/assembly array

UO2 TRISO ceramic coated particle

Fuel block length (m)1Number of fuel columns in core90

Average fuel enrichment

14%

Average fuel burnup (GWd/ton)

120

Fuel Cycle (months)

48

Number of safety trains

4

Emergency safety systems

inherent

Residual heat removal systems

inherent

Refueling Outage (days)

30

Distinguishing features

Multiple applications of power generation, hydrogen production,
process heat supply, steelmaking, desalination, district heating.

Modules per plant

Up to 4 reactors

Estimated construction schedule (months)

24-36

Seismic design>0.18 g automatic shutdown

Predicted core damage frequency per reactor year

<1 E-8/Design Status

Basic design225

5. Plant Safety Features
The reactor delivers fully inherent safety due to three enabling design features:
The ceramic coated particle fuel maintains its containment integrity under the designtemperature limit of 1600oC.
The reactor helium coolant is chemically inert and thus absent of explosive gas
generation or phase change.
The graphite-moderated reactor core is designed having characterises of negative
reactivity coefficient, low-power density and high thermal conductivity.

•••

As a result of these features, the decay heat of the reactor core can be removed by natural
draft air cooling from 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 will remain below the fuel design limit
as illustrated in Figure 6.
1. Ceramic (SiC) coatedfuel particle

3. Graphite core

Undergroundbuilding

Reactorpressure
vesselAir

Proven integrity at 1600oC

Air

Negative reactivity coefficient,high heat capacity and large
thermal conductivity of graphite
core provide for safe removal ofcore decay heat to external VCS.

Earth
Fuel kernel

50

TRISO ceramic coatings
Fuel pinNo failure offuel coatingat < 1600°C

0
1000

1600

2000

Temperature (oC)

Fuel block
3000

Experimental result

2. Inert helium coolant
No explosions of H2 and vapor due
to chemical inertness and absenceof phase change of helium coolant

3000

Core
Heat
radiation Heat
conduction

Vessel coolingsystem (VCS)

Reactor is safely shutdown andcooled by inherent design features
without reliance on any equipment
or operator action in the event ofloss of coolant or station blackout.

Fuel temperature (oC)

Failure fractionof coating (%)

100

2000 Fuel design limit 1600oC
1000
0Fuel temperature01

2 3 4 5 6
Elapsed time (day)7

Simulation of loss-of-coolant

Figure 6: Inherent reactor safety design (Reproduced courtesy of JAEA)
6. Plant Operational Performances
The ability to follow variable power and heat loads [5] is simulated as shown in Figure 7.
The simulation examines the plant response to an electric demand increase of 5% of the baserate per minute with corresponding reduction of the heat rate, which is the maximum
requirement for cogeneration load follow. The reactor remains at 100% power at all times.
Starting from the base cogeneration rates, the turbine power generation is increased to follow
the electric load demand increase by increasing the primary coolant inventory through theinventory control valve IV1. The IHX heat rate to the thermal production plant is lowered bylowering the intermediate loop flow circulation rate with the variable speed gas circulator. As
the primary exit temperature of the IHX begins to rise, the valve CV4 is opened, by active orprescheduled control to follow load demand, to direct cold flow from the compressor

226

1050

Reactor power

60

1000

Reactor outlet temperature

40

950

20
0

900

800

Turbine inlet

3

600
400

2

Cycle pressure ratio

200

1

0

0

0

1

2

3

4

5

6

Elapsed time (min)

7

8

200

Electricity generation rate

200

100

IHX heat rate

100

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)

4

Cycle pressure ratio

1000

IV1 flow rate (kg/s)

Reactor power (%)

80

300

300

Heat rate (MWt)100

Temperature (o C)

Temperature (o C)

1100

120

Electricity rate (MWe)discharge to mix with the hot exit gas of the IHX primary side. By applying flow bypass via
CV4 the goal to maintain the turbine inlet temperature near the rated 850oC is achieved, as
shown in Figure 8. The power sent out to external grid increases to 276 MW(e) from 178
MW(e) in as little as 7 minutes. The pressure in the reactor and at turbine inlet increases to 7
MPa from 5 MPa. To return to the base cogeneration state, the control is reversed by reducing
primary coolant inventory through another inventory control valve IV2 and simultaneously
by closing the bypass valve VC4.

8

Figure 7: Simulation of cogeneration plant load follow to +5%/min electric load increase(Reproduced courtesy of JAEA)
7. Instrumentation and Control System
Figure 8 illustrates the plant control system and instrumentation needed to enable load
following in power generation mode as well as in cogeneration mode. The overall approachto dynamic operation integrates the following four load control strategies:••••

Control of turbine speed, Sd, through flow bypass valve CV1.
Control of recuperator low-pressure-side inlet temperature, Tx, through flow bypass
valve CV2.
Control of turbine inlet temperature, Tt, by flow bypass valve CV3.
Control of turbine inlet temperature and pressure, Tt and Pt, by bypass valve CV4,and inventory flow valves IV1 and IV2.

The first two strategies are used to control rapid transients such as a sudden loss of electric
generator load. They are effective to protect the gas turbine from excess overspeed andprevent thermal shock in the recuperator.
The third strategy is used to automate heat rate to follow slow or fast changes in the heat loadof the IHX when it is perturbed by the thermal production plant. As the IHX primary exit
flow temperature rises or falls in response to a change in the IHX secondary heat load, theflow valve CV3 opens or closes to introduce more or less cold flow upstream of the turbine(from the compressor discharge to the turbine inlet) to keep the turbine inlet temperature
constant. The overall control strategy aims to continue normal power generation, unaffectedby any heat load change in the IHX.

227

Figure 8: Control system for GTHTR300 (Reproduced courtesy of JAEA)
The fourth control strategy is applied to automate cogeneration load follow. The conditions to
be met include (1) constant reactor temperature to avoid thermal stress in high temperature
structure; (2) constant reactor thermal power to yield base load economics; and (3) constant
power generation efficiency over a broad range of load follow.
8. Plant Design and Licensing Status
The design is developed at pre-licensing basic design stage. The design and development areplanned to be concluded to prepare for the lead plant construction around 2025.
[1]
[2]
[3]
[4]

[5]

228

X. Yan, K. Kunitomi, T. Nakada And S. Shiozawa, “GTHTR300 design anddevelopment”, Nucl. Eng. Des., 222, 247 (2003).
H. Sato, X. Yan, Y. Tachibana, K. Kunitomi, “GTHTR300—A nuclear power plantdesign with 50% generating efficiency”, Nucl. Eng. Des., 275, 190 (2014)
K. Kunitomi, X. Yan, T. Nishihara, N. Sakaba And T. Mouri, “JAEA’s VHTR forhydrogen and electricity cogeneration: GTHTR300C”, Nuclear Engineering andTechnology, 39, 1, 9 (2007).
S. Kasahara, N. Tanaka, H. Noguchi, J. Iwatsuki, H. Takegami, X. L. Yan And S.
Kubo, “JAEA’s R&D on the thermochemical hydrogen production IS process”,
HTR2014, Weihai, China, 27-31 Oct., Paper HTR2014-21233, Institute of Nuclear
and New Energy Technology (2014).
X. Yan, H. Sato, Y. Tachibana, K. Kunitomi, R. Hino, “Evaluation of High
Temperature Gas Reactor for Demanding Cogeneration Load Follow”, J. Nucl. Sci.
Technol., 49, 1, 121 (2012)

GT-MHR (OKBM Afrikantov, RussianFederation)
1. Introduction
The gas turbine modular helium reactor (GT-MHR) couples an HTGR with a Brayton power
conversion cycle to produce electricity at high efficiency. As the reactor unit is capable ofproducing high coolant outlet temperatures, the modular helium reactor system can also
efficiently produce hydrogen by high temperature electrolysis or thermochemical water
splitting.
The use of modular helium reactor units makes the system flexible and allows the possibility
to use various power unit schemes: with gas-turbine cycle, steam-turbine cycle and with thecircuit supplying high-temperature heat to industrial applications. The modular hightemperature gas-cooled reactor unit possess salient safety features with passive decay heatremoval providing a high level of safety even in case of total loss of primary coolant.
The modular helium reactor design proved the possibility of unit modularity with a wide
power range of a module (from 200 to 600 MW(th)) and NPP power variation as a functionof module number. This provides good manoeuvring characteristics of the reactor plant (RP)for regional power sources.

Figure 1: Reactor System Configuration of GT-MHR(Reproduced courtesy of OKBM Afrikantov)

229

2. Target Application
The GT-MHR can produce electricity at high efficiency (approximately 48%). As it iscapable of producing high coolant outlet temperatures, the modular helium reactor system
can also efficiently produce hydrogen by high temperature electrolysis or thermochemical
water splitting.
3. Development Milestones
1993

Minatom / General Atomics MOU on joint GT-MHR development for commercial units

1994

Russia proposes to build GT-MHR at Seversk to burn Russian WPu

1996

Framatome & Fuji Electric join the GT-MHR program

1997

Conceptual Design completed

1998

GT-MHR becomes an option within the US/RF Pu Disposition Strategy

1999

Conceptual Design review by international group of experts

2000

Work started on Preliminary Design

2002

Project review by Minatom of Russia and US DOE experts

2002

Reactor Plant Preliminary Design completed

2003

Begin Demonstration of key technologies

4. General Design Description
Design Philosophy
The GT-MHR direct Brayton cycle power conversion system contains a gas turbine, anelectric generator and gas compressors. The layout can be seen in Figure 1. 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,
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.
Reactor Core and Fuel Characteristics
Coated particle fuel is used. The fuel kernel (U or Pu oxide) is coated by a first porous layerof pyrocarbon, followed by a dense layer of pyrocarbon, a silicon carbide layer and an outerdense layer of pyrocarbon. Thousands of coated 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 for the commercial
GT–MHR utilizes low enriched uranium (LEU) in a once through mode without reprocessing
but alternative cycles including the disposition of plutonium were also studied in detail. The
GT-MHR show good proliferation resistance characteristics. It produces less total plutonium
and 239Pu (materials of proliferation concern) per unit of energy produced.

230

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

OKBM Afrikantov

Country of origin

Russian Federation

Reactor type

Modular helium reactor

Electrical capacity (MW(e))

285Thermal capacity (MW(th))

600

Expected capacity factor (%)

80

Design life (years)

60

2Plant footprint (m )9110

Coolant/moderator

HeliumPrimary circulation

Forced circulation

System pressure (MPa)7.24oCore inlet/exit temperatures ( C)

490 / 850

Main reactivity control mechanism

Rod insertion

RPV height (m)31

RPV diameter (m)10.2

RPV or module weight (metric ton)

N/A

Configuration of reactor coolant system

Loop/Integral type

Power conversion process

Direct Brayton or Indirect Rankine cycle

Fuel type/assembly array

Hexagonal Prism graphite block of 0.36m coated particle fuel

Fuel assembly active length (m)~0.8 m per block; 10 fuel blocks thus ~8 m active length

Number of fuel assemblies~1020 fuel blocks

Fuel enrichment (%)

LEU or WPu

Fuel burnup (GWd/ton)

100 – 720 (depending on the fuel cycle)

Fuel cycle (months)

25

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (active + passive)

Number of safety trains

2

Refuelling outage (days)

25

Distinguishing features
Modules per plant

Inherent safety characteristics; No core melt; High temperatureprocess heat capabilities; Small number of safety systems.
1 (depends on application)

Target construction duration (months)

48

Seismic design

8 points (MSK 64)

Core damage frequency (per reactor-year)

Core damage frequency not applicable to HTGRs
BDBA frequency < 1E-5 /year
Frequency of ultimate release at BDBA < 1 E-7 /year

Design Status

Preliminary design completed; Key technologies are being
demonstrated

231

The Power Conversion System Flow Diagram

Figure 2: GT-MHR flow diagram (Reproduced courtesy of OKBM Afrikantov)(1)-reactor, (2)-gas turbine, (3)-recuperator, (4), (6)-precooler and intercooler,(5), (7) - low-pressure and high-pressure compressors, (8) – generator, (9) – generator cooler,(10) – bypass valve, (11)….(15) – SCS components, (16), (21) – surface cooler of reactor
cavity cooling system (RCCS), (17) – air ducts, (18) – heat exchanger with heat pipes
The Brayton power conversion with direct gas turbine is adopted that contains a gas turbine,
an electric generator and gas compressors. This is shown in Figure 2. 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.
5. Safety Features
Safety is ensured by application of passive principles of system actuation. The decay andaccumulated heat is removed from the core through reactor vessel to reactor cavity coolingsystem 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. The design features, which determine theinherent safety of high-temperature reactor, are the following:
•

•
•

232

Using of helium coolant, which has some specific properties. During plant operation,
helium is not affected by phase transformations, does not dissociate, is not activatedand has good heat transfer properties. Helium is chemically 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 ensurescore configuration preservation under any accident;
Nuclear fuel in the form of coated fuel particles with multilayer ceramic coatings,
which retain integrity and effectively contain fission products under high fuel burnup•and high temperatures;
As uranium fuel is used, temperature and power reactivity coefficients are negativethat provides the reactor safety in any design and accident conditions.

6. Plant Arrangement

Figure 3: GT-MHR layout (Reproduced courtesy of OKBM Afrikantov)
7. Design and Licensing Status
Reactor Plant Preliminary Design completed with the demonstration of key technologies
underway.

233

MHR-T reactor/Hydrogen production complex(OKBM Afrikantov, Russian Federation)
1. Introduction
The MHR-T reactor/hydrogen production complex makes use of the basic GT-MHR reactor
unit design as the basis for a multi-module plant for hydrogen production. The hydrogenproduction through the steam methane reforming process or high-temperature solid oxide
electrochemical process is performed by coupled the plant with the modular helium
reactor(s). The use of modular helium reactor units makes the system flexible and allows thepossibility to use various power unit schemes: with gas-turbine cycle (the GT-MHR), 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 heat removal providing a high level of safety even in case of total
loss of primary coolant.

Figure 1: MHR-T RP layout – variant with steam methane reforming(Reproduced courtesy of OKBM Afrikantov)
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 coupledwith a modular helium reactor called MHR-T. The chemical-technological sector with steam
methane reforming is considered as an option for short-term perspective.

234

3. Development Milestones
2001

Pre-conceptual proposal

2005

Conceptual design completed

2007

Elaboration of technical requirements

4. General Design Description
Specific Design Features
The MHR-T complex includes the chemical-technological sector (hydrogen production
sector) and the infrastructure supporting its operation. The chemical-technological sector
includes hydrogen production process lines, as well as systems and facilities supporting theiroperation. The following processes are considered as the basic processes for the chemicaltechnological sector: (i) steam methane reforming; and (ii) high-temperature solid oxide
electrochemical process of hydrogen production from water. Heat shall be transferred directlyfrom primary coolant to chemical-technological sector medium in a high-temperature heatexchanger. The key component of chemical-technological sector medium circulating throughthe 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 based on:••••

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, highburnup and burial of fuel 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 ofheat 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 initialsteam-gas mixture is converted to hydrogen-enriched converted gas (mixture of water steam,СО, Н2, СО2, and СН4) in the course of a thermochemical reaction.
Reactor Core and Fuel Characteristics
Coated particle fuel is used. The fuel kernel (U oxide) is coated by a first porous layer ofpyrocarbon, followed by a dense layer of pyrocarbon, a silicon carbide layer and an outerdense layer of pyrocarbon. Thousands of coated 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 lowenriched uranium (LEU) in a once through mode. The MHR-T show good proliferation
resistance characteristics. It produces less total plutonium and 239Pu (materials of proliferation
concern) per unit of energy produced. The fuel form presents formidable challenges todiversion of materials for weapons production, as either fresh or as spent fuel.

235

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

OKBM Afrikantov

Country of origin

Russian Federation

Reactor type

Modular helium high-temperature reactor

Electrical capacity (MW(e))

4х205.5

Thermal capacity (MW(th))

4x600

Expected capacity factor (%)

80

Hydrogen or MHM production efficiency (gross), %80

Hydrogen production (t/h)

4x12.5

Design life (years)

60

2Plant footprint (m )N/ACoolant/moderator

Helium/graphite

Primary circulation

Forced circulation

System pressure (MPa)7.5oCore inlet/exit temperatures ( C)

578 / 950

Main reactivity control mechanism

Rod insertion

RPV height (m)32.8

RPV diameter (m)6.9

RPV or module weight (metric ton)

N/A

Configuration of reactor coolant system

IntegralPower conversion process

Direct closed Brayton cycle

Fuel type/assembly array

Hexagonal Prism graphite block of 0.36m coated particle fuel

Fuel assembly active length (m)~0.8 m per block; 10 fuel blocks thus ~8 m active length

Number of fuel assemblies~1020 fuel blocks

Fuel enrichment (%)

20

Fuel burnup (GWd/ton)

125

Fuel cycle (months)

30

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (active + passive)

Number of safety trains

2

Refuelling outage (days)

25

Distinguishing features
Modules per plant

Multi-module HTGR dedicated to hydrogen production / hightemperature process heat application.
4

Target construction duration (months)

48

Seismic design

8 points (MSK 64)

Core damage frequency (per reactor-year)

Core damage frequency not applicable to HTGRs
BDBA frequency <1E-5/year
Frequency of ultimate release at BDBA <1E-7 /year

Design Status

Conceptual design236

Power Conversion System
A power conversion unit is integrated in a single vessel and includes a vertical turbomachine,
highly efficient plate heat exchanger, and coolers. A high-temperature heat exchanger (IHX)
for the MHR-T option with methane steam reforming is an integral part of the thermal
conversion unit and is partitioned as a three-stage heat exchanger. Arrangement of the heatexchanger 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 aseparate heat exchanger consisting of several modules. The material of the heat exchange
surface of the module is a heat-resistant alloy.

Figure 2a: Power conversion unit

Figure 2b: High temperature heat exchanger section(Reproduced courtesy of OKBM Afrikantov)
The turbomachine includes a generator and turbo-compressor mounted on a single shaft onelectromagnetic suspension. Gas turbine cycle of power conversion with the helium
turbomachine, heat exchanger and intercooler provides thermal efficiency at 48%.
Reactor Coolant System
Working media in circulation circuits are helium of the first circuit and steam-gas mixture(SGM) in the CTS. The peculiarity of heat exchangers for the production of hydrogen bymethane reforming process is the transfer of heat from high temperature helium of the
primary circuit to the chemically aggressive medium of hydrogen production circuit.
5. Safety Features
The safety features of the MHR-T reactor are the same as for the GT-MHR and it is thus
repeated below. In addition special attention is given to external shock waves (due to an
explosion at the chemical plant). Safety is ensured by application of passive principles ofsystem actuation. The design features, which determine the inherent safety of hightemperature reactor, are the following:•

Using of helium coolant, which has some specific properties. During plant operation,
237

•
•
•

helium is not affected by phase transformations, does not dissociate, is not activatedand has good heat transfer properties. Helium is chemically 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 ensurescore configuration preservation under any accident;
Nuclear fuel in the form of coated fuel particles with multilayer ceramic coatings,
which retain integrity and effectively contain fission products under high fuel burnup
and high temperatures;
As uranium fuel is used, temperature and power reactivity coefficients are negativethat provides the reactor safety in any design and accident conditions.

Decay heat removal system
The decay and accumulated heat is removed 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.
6. Plant Arrangement
The main components of each NPP module are arranged in isolated premises of theunderground containment of the NPP main building. The chemical-technological sector
equipment is arranged outside the containment of the NPP main building. The MHR-T
energy-technological complex is designed for a specific site on the basis of design solutions
selected with account of:••

Climatic conditions typical of central Russia;
Special external impacts – 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 more than one MHR-T module. The main
reactor equipment is arranged in a vertical vessel located in a separate cavity parallel to thepower conversion unit and high-temperature heat exchanger vessels.

Figure 3: MHR-T site with hydrogen production complex(Reproduced courtesy of OKBM Afrikantov)
7. Design and Licensing Status
Conceptual Design completed.

238

MHR-100 (OKBM Afrikantov, RussianFederation)
1. Introduction
The design basis for these design elaborations is the world-wide experience in thedevelopment of experimental HTGR plants. Russia has experience (more than 40 years) inthe development of HTGR plants of various power (from 100 to 1000 MW) and for various
purposes, has established the experimental facilities for the R&D work, the fuel element andmaterial fabrication technology, including the fabrication and mastering of pilot equipment,
and various activities in hydrogen generation technology. Today, stations with electric
capacity ~300 MW(th), which are placed on all territory of Russia and adapted to regional
systems, provide need of Russia in the electric power and are actually the regional powerindustry. The regional power industry consists mainly of cogeneration plants producing about
40% of electric power and 85% of the heat generated in Russia. Analysis shows that small
and medium NPP with HTGR have therefore good prospects to add to or replace theseregional 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 predicted study of the power market development anddemands, Russia has established Rosatom enterprise cooperation and performed preconceptual developmental work for commercial small MHR100 RP prototype with modular
helium reactor and several power conversion layouts as sources of various power-industrial
applications. Within the developmental work the following MHR100 options for industrial
applications were studied:••••Electric power and district heat production by core thermal power conversion toelectric one in direct gas-turbine Brayton cycle – MHR-100 GT;
Electric power and hydrogen generation by high-temperature steam electrolysis
method – MHR-100 SE;
Hydrogen generation by steam methane reforming method MHR-100 SMR;
High-temperature heat supply to oil refinery plant – MHR-100 OR.

2. Target Application
The MHR-100 is intended for regional power generation and heat production in the RussianFederation. A single reactor unit design can be implemented in various plant configurations.
3. Development Milestones
2014

Conceptual design completed

4. General Design Description
Specific Design Features
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 (215MW(th)) was selected according to: (i) the regional power industry and district heat supply
needs; (ii) the manufacture needs in high- and medium-temperature heat supply fortechnological processes; and (iii) process capabilities of national enterprises in fabrication of
RP main components including vessels.

239

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

OKBM Afrikantov

Country of origin

Russian Federation

Reactor type

Modular helium reactor

Electrical capacity (MW(e))

25 – 87 (depending on the configuration)

Thermal capacity (MW(th))

215

Expected capacity factor (%)

N/A

Design life (years)

N/A

2

Plant footprint (m )N/ACoolant/moderator

Helium / Graphite

Primary circulation

Forced circulation

System pressure (MPa)4–5oCore inlet/exit temperatures ( C)490-553 / 795 – 950

Main reactivity control mechanism

Rod insertion

RPV height (m)N/A

RPV diameter (m)N/A

RPV or module weight (metric ton)

N/A

Configuration of reactor coolant system

Loop type

Power conversion process

Direct cycle Brayton, cogeneration / IHX

Fuel type/assembly array

Hexagonal Prism graphite block with coated particle fuel

Fuel assembly active length (m)N/A

Number of fuel assemblies

N/A

Fuel enrichment (%)

LEU < 20% enriched

Fuel burnup (GWd/ton)

N/A

Fuel cycle (months)

N/A

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (passive + active)

Number of safety trains

2

Refuelling outage (days)

25

Distinguishing features

A multipurpose reactor for electricity production, heat generation,and hydrogen generation; high-temperature heat supply to oil
refinery plant

Modules per plant

Depending on deployment

Target construction duration (months)

48

Seismic design

8 points (MSK 64)

Core damage frequency (per reactor-year)

Core damage frequency not applicable to HTGRs
BDBA frequency < 1E-5 /year
Frequency of ultimate release at BDBA < 1E-7 /year

Design Status

Conceptual design

5. Safety Features
Safety is ensured by application of passive principles of system actuation. The decay andaccumulated heat is removed from the core through reactor vessel to reactor cavity coolingsystem 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
240

circulation systems and power sources failure. See other Russian design modular HTGR for
more information.
6. Design Variants based on the modular MHR-100

Figure 1: MHR-100 GT layout/options(Reproduced courtesy of OKBM Afrikantov)
Major Technical Parameters of MHR-100 GT
Parameter

Power Mode

Cogeneration mode

Reactor heat capacity (MW)

215

215

Net power generation efficiency (%)

46.1

25.4

Helium temperature at reactor inlet/outlet (°C)

558/850490/795

Low-pressure helium temperature at recuperator
inlet (°C)

583

595

Helium flow rate through the reactor (kg/s)

139.1

134

Helium bypass flow rate from HPC outlet torecuperator outlet at high-pressure side (kg/s)

–

32.2

Helium pressure at reactor inlet (MPa)

4.91

4.93

Expansion ratio in turbine

2.09

1.77

Generator/TC rotation speed (rpm)3000/9000

3000/9000

PCU cooling water flow rate (kg/s)

804

480

Delivery water temperature at inlet/outlet (°C)–

70/145The 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 coated fuel particles (based on UO2),high burnup and possibility to disposal the spent fuel blocks without additional
reprocessing;High-performance high-temperature and compact heat exchangers, high-strength
casings of heat-resistant steel;
241

•
•
•

Direct gas-turbine cycle of power conversion with high-efficiency recuperation andintermediate coolant cooling;
Experience in high-efficiency gas turbines application in power engineering andtransport;
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 ofpower and process parts. The power part is unified to the maximum for all options and is a
power unit consisting of a reactor unit with a thermal power of 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 forhigh-temperature heat supply to various technological processes, depending on the purpose.

Figure 2: MHR-100 SE layout/option (Reproduced courtesy of OKBM Afrikantov)
Major Technical Parameters of MHR-100 SE
Parameters

ValuesReactor heat capacity (MW)215

Useful electric power of generator (MW)87.1

Net power generation efficiency (%)

45.7

Helium temperature at reactor inlet/outlet (°C)

553/850

Helium flow rate through the reactor (kg/s)

138

Helium pressure at reactor inlet (MPa)

4.41

Expansion ratio in turbine

2.09

Generator/TC rotation speed (rpm)Helium flow rate through the turbine (kg/s)

126

Helium temperature at PCU inlet/outlet (°C)

850/558

SG power (MW)22.3

Helium flow rate through SG (kg/s)

12.1

Helium temperature at SG inlet/outlet (°C)

242

3000/9000

850/494

Steam capacity (kg/c)

6.46

Steam pressure at SG outlet (MPa)

4.82

The unified gas-turbine PCU is planned to be used in MHR-100 GT and MHR-100 SE
options. Vertical oriented TM is the main feature of the PCU and consists of the turbocompressor (TC) and generator with rotors, 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 pre-cooler andintercooler 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 precooler and intercooler by the cooling water system, then in dry fan cooling towers toatmospheric air.

Figure 3: MHR-100 SMR layout/options (Reproduced courtesy of OKBM Afrikantov)
Major Technical Parameters of MHR-100 SMR
Parameters

ValuesReactor heat capacity (MW)215

Helium temperature at reactor inlet/outlet (°C)

450/950

Helium flow rate through the reactor (kg/s)

81.7

Helium pressure at reactor inlet (MPa)5.0Steam-gas mixture pressure at HX inlet (MPa)

5.3

HX-TCF 1
HX 1 capacity (MW)31.8

Helium/steam-gas mixture flow rate (kg/s)

12.1/43.5

Steam-gas mixture temp. at inlet/outlet (°C)

350/650

HX-TCF 2
58.5

HX 2 capacity (MW)Helium/steam-gas mixture flow rate (kg/s)

22.2/60.9

Steam-gas mixture temp. at inlet/outlet (°C)

350/750

HX-TCF 3
HX 3 capacity (MW)125

Helium/steam-gas mixture flow rate (kg/s)

47.4/101

Steam-gas mixture temp. at inlet/outlet (°C)

350/870

243

Heat exchange blocks are intended to transfer heat power from the reactor to the consumer of
power-technological applications. Depending on the working fluid, process type andprobability of radioactivity ingress to the process product and contamination of equipmentwith radioactive products, two- or three-circuit RP configuration can be used. So, two circuit
configurations are used in MHR-100 SE NPP for hydrogen generation and in MHR-100 SMR
for steam methane reforming. Water steam is the main component of process fluid in theseprocesses. The analysis shows that the effects of hydrogen-bearing products ingress inpotential accidents with depressurization of the steam generator (SG) or high-temperature
heat exchanger (HX) are reliably checked by reactor control and protection systems. MHR100 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-temperature intermediate helium-helium HX (IHX) and intermediate helium circuitand then to network circuit of petrochemical applications. This decision restricts radioactivity
release to the network circuit and provides radiological purity of the process product andminimum contamination of the primary circuit with process impurities.

Figure 4: MHR-100 OR layout/options (Reproduced courtesy of OKBM Afrikantov)
Major Technical Parameters of MHR-100 OR
Parameters
Reactor heat capacity (MW)244

Values215

Helium temperature at reactor inlet/outlet (°C)

300/750

Helium flow rate through the reactor (kg/s)

91.5

Helium pressure at reactor inlet (MPa)5.0

IHX capacity (MW)217

Primary/secondary helium flow rate through IHX (kg/s)

91.5/113

Primary helium temp. at IHX inlet/outlet (°C)

750/294

Secondary helium temp. at IHX inlet/outlet (°C)

230/600

Secondary helium pressure at IHX inlet (MPa)

5.50

PBMR-400 (Pebble Bed Modular Reactor
SOC Ltd., South Africa)
1. Introduction
The Pebble Bed Modular Reactor is a High Temperature Gas Cooled Reactor based on theevolutionary design of the German HTR-Module design. The PBMR is designed in a modular
fashion to allow for additional modules to be added in accordance with 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 underdevelopment since 1996. Most of these designs are based on a direct Brayton cycle as thisholds promise of higher efficiencies. The maximum achievable power levels for the reactor
was increased in several design steps in order to reach a set target for installed cost/kW thatwould be roughly comparable to coal fired power when lifetime costs were evaluated. As aresult the design of the reactor core evolved from the original base of 200 MW(th) adopted
from the HTR-Module design to reach 400 MW(th) with an annular core, as shown in
Figure 1. As a result of the world financial crisis in 2008 and short term funding constraints arethink of the product priorities lead to a decision to concentrate on the electricity and process
heat market with a single reactor product and thus a decision was made to use an indirect
steam cycle. The direct cycle design was archived with a view to further progress this design
when conditions (financial and technology development in materials for the direct cycle)
improve.

Figure 1: Reactor System Configuration of PBMR-400(Reproduced courtesy of Pebble Bed Modular Reactor SOC Ltd.)245

2. Target Application
The PBMR-400 can produce electricity at high efficiency via a direct Brayton cycle
employing a helium gas turbine.
3. Development Milestones
1993

The South African utility Eskom identifies PBMR as an option for new generating capacity

1995

Start of the first pre-feasibility study

1999

Design optimization: PBMR-268 with dynamic central column

2002

Design changed to PBMR-400 with fixed central column

2002
2004

The Pebble Bed Micro Model (PBMM) demonstrated the operation of a closed, three shaft, pre- and intercooled Brayton cycle with a recuperator.
Vertical layout of turbo machines changed to conventional single horizontal layout

2006

Commissioning of Helium Test Facility for full scale system and component tests

2006

Tests starts in the Heat Transfer Test Facility

2007

Advanced fuel coater facility commissioned

2009

Coated particles send for irradiation testing at INL

2009

Financial crises has funding implications and alternative process heat markets and designs are studied

2010

Project closure

2014

Project in care and maintenance

4. General Design Description
Design Philosophy
The PBMR-400 is a high-temperature helium-cooled, graphite moderated pebble bed reactor
with a multi-pass fuelling scheme. The design safety targets and features means that thereactor can be deployed close to the end 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 shelteror be evacuated. To achieve this goal there shall be no need for engineered or moving
mechanical components to ensure this target is met while the exposure to plant personnel
shall also be significantly lower than today’s best international practice.
The Brayton Cycle Power Conversion System
The Brayton power conversion with direct gas turbine is adopted. It is a closed cycle where
the helium coolant is used to transport heat directly from the core to power turbine. The
design incorporates a single shaft for the turbine, the compressors and the power generator. A
recuperator is included. The layout is shown in Figure 2 that also shows the pre- and intercooler as well as other auxiliary systems. From the reactor unit the hot helium enters directly
to the turbine where energy is used to drive the shaft and therefore the electric generator andcompressors. From the turbine the helium then passes consecutively through the low-pressure
primary side of the recuperator, then the pre-cooler, the low pressure compressor, intercooler,high pressure compressor and then on to the high-pressure secondary side of the recuperator
before re-entering the reactor unit. The direct gas cycle is attractive since it promises thebenefits of simplification, with the potential of lowering the capital and operational costs.
Due to the high outlet gas temperatures one will also expect a substantial increase in theoverall system efficiency.

246

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

PBMR SOC LtdCountry of origin

South Africa

Reactor type

Modular High Temperature Gas-cooled reactor

Electrical capacity (MW(e))

165

Thermal capacity (MW(th))

400

Expected capacity factor (%)

93

Design life (years)

40

2Plant footprint (m )4200 (main structures only)

Coolant/moderator

Helium / Graphite

Primary circulation

Forced circulation

System pressure (MPa)9oCore inlet/exit temperatures ( C)

500 / 900

Main reactivity control mechanism

Negative temperature coefficient; Control rod insertion

RPV height (m)30

RPV diameter (m)6.2 (inner)

RPV or module weight (metric ton)

N/A

Configuration of reactor coolant system

Loop type

Power conversion process

Direct Brayton cycle

Fuel type/assembly array

Pebble bed with coated particle fuel

Fuel pebble diameter (cm)6

Number of fuel assemblies~452000 in core

Fuel enrichment (%)

9.6

Fuel burnup (GWd/ton)

92 (average discharge)

Fuel cycle (months)

N/A; Online / on-power refuelling

Cogeneration capability

No

Approach to engineered safety systems

ActiveNumber of safety trains

N/A; The reactor cavity cooling system is not a safety system but
only for investment protection. It is divided in three trains,
oversized with active / passive operational modes.

Refuelling outage (days)

N/A since online; Major outage for maintenance every 6 years

Distinguishing features

Inherent safety characteristics; No core melt; High efficiency;
Small number of safety systems

Modules per plant

1 (could also be in multi-module plant)

Target construction duration (months)

36 months (for the nth unit)

Seismic design

0.4g PGA for main power system designCore damage frequency (per reactor-year)

Core damage frequency not applicable to HTGRs
No off-site shelter or evacuation

Design Status

Preliminary design completed; Test facilities demonstration;
Project stopped in 2010 and design in care and maintenance247

Figure 2: PBMR-400 power conversion unit components and layout(Reproduced courtesy of Pebble Bed Modular Reactor SOC Ltd.)The reactor power is adjusted by changes in the helium the mass flow rate in the power
conversion unit. Since pressure changes will affect the helium density it is one mechanism toachieve this since the mass of the coolant will change accordingly. The helium inventory
system is used and power control is performed in combination with bypass valves. Higher
mass flows will remove more heat from the core eventually leading to an increase in thefission power through the natural feedback effects. Power reduction is achieved by removing
gas from the circuit.
Reactor Core and Fuel Characteristics
The core neutronic design results in 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 coreheight is 11 m. In steady state (equilibrium core) operation the fuel sphere powers (maximum
2.7 kW) and operational temperatures (<1100 oC) fulfil the design criteria. The core contains~ 452000 fuel spheres or “pebbles” with a packing fraction of 0.61.

Figure 3: PBMR-400 coated fuel particle(Reproduced courtesy of Pebble Bed Modular Reactor SOC Ltd.)

248

The fuelling scheme employed is the continuous on-line multi-pass method. Fresh fuelelements are added to the top of the reactor while used fuel pebbles are removed at thebottom to keep the reactor at full power. On average fuel is circulated six times through thereactor. This reduces power peaking and maximum fuel temperatures in normal operation andloss of coolant conditions. Coated particle pebble fuel is used as shown in Figure 3. The fuelkernel (UO2) is coated by a first porous layer of pyrocarbon, followed by a dense layer ofpyrocarbon, a silicon carbide layer and an outer dense layer of pyrocarbon. About 15000
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 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 are provided by two
separate and diverse systems while the overall negative reactivity temperature coefficient isnegative over the total operational range. The reactivity bound by the control rods allow
xenon override for load follow between 40% and 100%.
Heat Removal
The Reactor Cavity Cooling System (RCCS) is a means to remove residual heat passively fora defined time, and indefinitely with the use of an active system after refilling the coolingsystem. For this to work, the RPV 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 cooling by the main circulation system, thecavity cooler and/or the building structural materials are able to limit the increase in fueltemperature in the most affected region of the core to below the allowable fuel temperature
limit.
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. It’s 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 humanintervention 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 a decreasing curve without any core failure or release of radioactivity to theenvironment. The PBMR design has the following attributes that contribute to enhanced
safety:•••••Use of Triple Coated Isotropic (TRISO) fuel particles shown to remain intact to atleast 1600 °C and with some time delayed failure fractions at even higher
temperatures;
A geometry that allows the passive dissipation of decay heat to an external heat sink;
Relatively low power density to aid in the limitation of fuel temperatures following aloss of coolant due to an un-isolatable leak;
Load following limited to 100-40-100% to reduce the excess operating reactivity to avalue that prevents fuel failure for any scenario of group control rod withdrawal
without scram;Use of helium as coolant which avoids the effects of phase changes and has a

249

•
•
•

negligible effect on plant reactivity when pressures fluctuate;
Control rods move only in the reflector and thus avoid any problem of mechanicaldamage to the fuel spheres;
Optimize the heavy metal fuel loading (water ingress is just about eliminated in the
direct gas cycle power conversion design) - cannot cause undue reactivity addition;
Use of nuclear grade graphite to ensure minimal corrosion by impurities and low
activation at end of life.

Barriers and Containment approach
The most important barriers to fission product release are the coatings of the fuel particles.
High manufacturing requirements and strict quality control will ensure that the amount of
free fissile or fertile material in the graphite matrix will be very low, and the percentage offuel particles with missing or failed coatings will meet the requirements. This will ensure a
low activity level in the primary circuit. A second barrier is provided by the Helium Pressure
Boundary. This enclosure assures that only very small amounts of helium and nuclides
circulating within it are released into the building during normal operation. A third barrier isthe confinement building. This is that part of the reactor building which houses the primarycoolant pressure boundary. The vented confinement is designed for very low leakage at low
pressure, and to prevent damage to components important to safety, as well as to contain thebuild-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 withthe released gas filtered as required.
6. Plant Arrangement

Figure 4: PBMR-400 building layout(Reproduced courtesy of Pebble Bed Modular Reactor SOC Ltd.)7. Design and Licensing Status
Reactor Plant Preliminary Design completed and demonstration of key technologies were
underway when the project was terminated in 2010.

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HTMR-100 SMR (Steenkampskraal Thorium
Limited (STL), South Africa
1. Introduction
The HTMR-100 (High Temperature Modular Reactor) pebble bed reactor is a hightemperature 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 steamturbine/generator system to produce 35MW of electric power. The steam can be used in awide range of cogeneration applications. The reactor is also suitable to provide direct hightemperature energy for process heat. The design of the reactor is based on proven technology
and therefore no new basic technology development is needed. The size of the reactor and thefuel cycle were chosen to simplify the design and operation of the module. The approach tosmall intrinsic safe modular units ensures continuous production, easy road transportability,
skid mounted sub systems, wider range of manufactures, fast construction and an enhanced
licensing process.

CRDMs

ReactorCore
BarrelGraphite
Structures

Connecting
Vessel Unit
Core
Support
Structure

Figure 1: Reactor Configuration of the HTMR-100 (Reproduced courtesy of STL)
2. Target Applications
The HTMR-100 is capable of supplying 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 temperature process heat applications. The HTMR-100 is aperfect fit for clients who want to progressively extend their generating capability. The
unique safety characteristics make it possible to introduce and construct these plants to nonnuclear countries. First-world countries that want to utilize their stock of Plutonium forpeaceful applications are also markets for HTMR-100 reactors.
251

3. Development Milestones
2012

Project started

2017

Preparation for Pre-license application

2018

Conceptual design finished

4. General Design Description
Design Philosophy
The reactor has good load following characteristics which is needed for stand-alone (not gridcoupled) applications. The “once through then out” (OTTO) fueling scheme leads to a simple
and cost effective fuel management system. The relative low primary loop pressure requires athinner walled pressure vessel and thus an easier manufacturing process, resulting in a wider
range of vessel manufacturers. The HTMR-100 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.
Reactor Core and Power Conversion Unit
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 fordifferential expansion and volumetric changes due to temperature 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.
The power conversion system uses a helical coil steam-generator unit supplying super heatedsteam to the turbine. The main system will be supplied as four skid mounted units namely thecondenser, turbine, gearbox and electric generator. The turbine can be used in a back pressure
configuration or intermediate temperature steam can be taken off for process heatapplications.
PRESSURE RELIEVESYSTEMSub atmospheric
HVAC

Air
Passivereactor cavity
co olin g system

Service hall

Fro m and to
purified h elium
storage tanks

HELIUM INVENTORY
AND PRESSURE
CONTROL SYSTEM

Secured reactor
cavity coo lingsystem

Th100

PACS

TO ATMOSPHERE

SG

Steam generator
pressure relievesystemReactor building
annex(Water Tracts)

To Spent
Reactor cavity

Steam generator
cavityP

POWER CONVERSIONSYSTEM

Figure 2: Power conversion and main support systems (Reproduced courtesy of STL)

252

MAJOR TECHNICAL PARAMETERS:Parameter
Technology Developer

Value
Steenkampskraal Thorium Limited (STL)Country of origin

South Africa

Reactor typeHigh temperature Gas cooled Reactor (HTGR)Electrical capacity (MW(e))

35 for single module, 140 for four module plant

Thermal capacity (MW(th))

100 for single module

Expected availability Factor

˃95%

Design Life (years)

40 full power years

2

Plant Footprint (m )5000 (Buildings only)

Coolant/moderator

Helium as coolant, graphite as moderator

Primary circulation

Forced He circulation

System pressure

4 MPa

Main Reactivity Control Mechanism

Absorber Rods in the reflector

RPV Height (m)15.3

RPV Diameter (m)5.9 on flange

Coolant Temperature, Core Outlet (°C)

750

Coolant Temperature, Core Inlet (°C)

250

Integral Design

YesPower Conversion Process

Indirect Rankine Cycle

High-Temp Process Heat

Yes

Low-Temp Process Heat

Yes

Co-Generation Capability

Yes

Design Configured for Process Heat Applications

Yes

Passive Safety Features

Yes

Active Safety Features

Only Residual Heat Removal backed up by passive systems

Fuel TypeTRISO particles in pebbles: LEU, Th/LEU, Th/HEU or Th/Pu

Fuel size

60 mm diameter

Number of Fuel units~150000 pebbles ; about 125 pebbles/day throughput

Fuel Enrichment

Various – see fuel description

Fuel Burnup (GWd/ton HM)

80 to 90

Fuel Cycle

Continuous online fuel loading

Number of safety trains

No engineered safety train

Emergency active safety functions

None required

Residual heat removal systems

Passive and Active

Maintenance schedule ( frequency and outage(days)

Every 6 years/30 days, 12 years/50days, mid-life 180days

Distinguishing features

Modules per Plant

No core meltdown, modular design , reduced construction time, noactive ES systems, high quality steam , reactor upgradable to very
high temperature, spent fuel in highly acceptable form, minimal
production of tritium thus ideal for inland sites.
1 or multiples of 2 modules

Estimated Construction Schedule (months)

36 months , based on continuous manufacturing and construction

Seismic design

0.3 g for generic site. ( 0.5 g under consideration)

Core damage frequency

Slight damage with water ingress event with design base frequency< 1E-4 /year
Advanced Concept Phase

Design Status

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Fuel characteristics and Supply
The fuel elements (FE) for the HTMR-100 are 60 mm diameter spheres consisting of aspherical fuel zone of approximately 50 mm diameter, in which the TRISO-coated particles
are randomly distributed in the graphitic 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 HTMR-100 reactor, ranging from LEU
UO2 to mixtures of Th/LEU, Th/HEU and Th/Pu. The following options have already beenstudied and show to be viable:i.
LEU : <10% Enrichment, (7-10g HM/sphere)ii.
Th/LEU: various options
a. 50% LEU by mass 20% enrichment, 50% Th (10-12g HM/sphere)b. 25% LEU by mass 20% enrichment, 75% Th (16-20g HM/sphere)iii. Th/HEU, 10% HEU by mass 93% enrichment, 90% Th by mass (10-12g HM/sphere)iv.
Th/Pu: 15% reactor grade Pu by mass, (12g HM/sphere)
A Fuel Qualification and Test programme will be conducted on the fuel prior to loading of thereactor. The HTMR-100 operates on a much longer burn-up fuel cycle compared toconventional reactors. The non-proliferation 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
The Fuel Handling and Storage System (FHSS) is a support system to the Main Power
System of the HTMR-100.

Figure 3: Fuel handling scheme (Reproduced courtesy of STL)
A six-month supply of fresh fuel is kept in the fresh fuel storage facility. New spherical fuelelements (fresh fuel) 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 isopened 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.

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Spent fuel is removed from reactor pressure vessel by means of two fuel unloading machines(FUMs), connected to a single reactor defueling chute. Each FUM discharges spent fuel intoa discharge lock. The discharge locks have sufficient capacity for a full power day of spent
fuel elements.
Spent fuel, damaged in such a manner that gravity conveyance is compromised, ismechanically separated inside the FUM and accumulated in a dedicated scrap canister,
located inside the FUM.
Provision is made in the design of the FHSS for the temporary fitment of activity measuring
devices. These are to be used for the removal of the first core graphite spheres. Separate burnup measuring devices can also be temporarily installed for core performance andcalculational purposes when running fuel composition trials.
During decommissioning the FHSS extracts the last used fuel in the same manner as during
normal operation.
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 subgroups, 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 thecircular channels. Each rod is equipped with a position indicator which measures the positionof the rod over its entire positioning range and with position indicators for the upper andlower limit positions.
Reactor Pressure Vessel and Internals
The Reactor Pressure Vessel (RPV) is constructed to the ASME III subsection NB code. It
comprises of two main components:• reactor vessel body• vessel head which is bolted to the 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 ofthese 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 the neutron source. The functional design of thestructural core internals is such that they are capable of withstanding the steady state andtransient 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 fuel element flow.
The loads borne by the ceramic internals are transferred to the steel core barrel and then to
the reactor pressure vessel through metallic components such as the lower support structure
and the core barrel axial and radial supports.

255

All areas of the core internals are designed for the service life of the reactor. Access forceramic structure inspections can be done through the fuel loading channel and the reflector
rod holes.
5. Plant Safety Features
Engineered Safety System Configuration and Approach
In principle the plant is designed to perform its safety functions without reliance on theautomated 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. The low power density, the large mass of the core structures, the slender coregeometry and the use of materials capable of withstanding high temperatures ensure complete
passive residual heat removal capability without exceeding design limits of components.
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 withtemperature will shut down the reactor and maintain a condition where no damage to the fuel,core structures and reactor vessel occurs. Excessive reactivity increase during water or water
vapor ingress (increasing moderation) is prevented by designing the reactor for limited heavy
metal content of the fuel.
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 theinside of in the reactor cavity wall. Water is circulated through the tubes to form a cold wall.
The RCCS is a passive system and consists of three independent cooling trains and isdesigned for all postulated design basis conditions.
Containment Function
The primary fission product barrier is the TRISO coated fuel particles, which keep the fission
products contained under all postulated events, even if the second barrier (the primarypressure vessel system) and the third barrier (the building filter system) collapse.
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 thereactor, the preference is to change the ratio between steam supply and electricity supply.
Excess steam and/or electricity can be utilized in the desalination plants to provide water as asellable commodity earning additional revenue. This allows the plant to operate virtuallycontinually at full power very close to the plant availability. In order to maintain the plantsafety and operational performance, specific areas are controlled for human occupation andprotection.
7. Instrumentation and Control Systems
The Automation System (ATS) comprises that group of safety and non-safety I&C systems
that provide automated protection, control, monitoring and human-system interfaces. The
following three specific systems in the HTMR-100 system structure define control andinstrumentation:• Plant control, data and instrumentation system: Provides overall plant control andmonitoring. It includes architecture which combines I&C systems and the control

256

•
•

room equipment.
Equipment/Investment Protection System: Reduce the risk of losing critical or
expensive components by maintaining the plant within its normal and safe operatingenvelope.
Protection System: Initiates reactor SCRAM to protect against nuclear control failure
or loss of primary coolant, etc.

The HTMR-100 ATS architecture implements modern digital I&C designs and equipment.
The architecture of the overall Automation System is based on redundancy, diversity,
separation and functional and physical independence. This concept includes four levels ofdefence-in-depth defence: (1) Distributed Control System or Control System; (2) Reactor
Protection System; (3) Manual Shutdown System; and (4) Post-event Information System.
8. Site and Plant Layout
The HTMR-100 is configured to simplify the construction of the various buildings and toensure easy installation of the reactor, steam generator and handling of spent fuel. The
proposed plant layout aims to protect safety important functions while simplifying
operational and maintenance tasks. The plant layout allows for the addition of multiple
reactors (and associated equipment) in a compact multi-module fashion.
Reactor building
The reactor building contains the safety equipment that provides the necessary functions forthe safe shutdown of the reactor under all design basis conditions. The reactor building ispartially submerged below ground level such that the reactor and steam generator cavities arecompletely protected against postulated external threats. The depth can be further adapted tosuit the geological conditions of the specific site to provide for the necessary level of seismic
protection. The reactor building, electrical building and auxiliary buildings are connected bymeans of underground tunnels, providing protection for interlinked services and, it also
ensures that spent fuel is never brought above ground level. Provision is made for the storageof all spent fuel produced during the operating life of the plant. The reactor building isseismically designed to withstand a design basis earthquake (DBE) and together with thespent fuel storage bunker, is the only safety related building structure of the HTMR-100.

Figure 4: HTMR-100 Site and Plant Concept Layout (Reproduced courtesy of STL)

257

Electrical building
The EB 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 the control room.
Turbine Generator Building
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 safetyvalves.
Balance of Plant
The balance of plant consists of the secondary systems and the tertiary systems. The Power
Conversion System (PCS) has one set of isolation valves between itself and the Nuclear
Steam Supply System (NSSS). The PCS and its major subsystems are of non-nuclear safetyclass thus conventional safety class and the product is of high quality industrial standard.
Electric Power Systems
The Electrical Power Supply System (EPSS) for the single module HTMR-100 plant supplies
power to the plant safety and non-safety equipment for normal plant operation, start-up,
normal shutdown as well as for accident mitigation and safe shutdown. The EPSS provides
the following voltage levels: 230VAC, 400VAC, 11kVAC, 24/48VDC and 250VDC. The
primary voltage of the Station Feeder Transformer is transmission grid dependent and thus
site specific. The majority of plant loads are powered from the Main Power Supply System(MPSS) while all safety-related loads as well as non-safety related loads on the Nuclear
Island (NI) portion of the EPSS are powered by the Reactor Power Supply System (RPSS).
The Reactor Protection System (RPS) and some non-safety related loads (i.e. Primary
Blowers 1 & 2, etc.) are powered from the RPSS. A portion of the RPSS circuit (inside thereactor building) is categorized as safety-related (Class1E).
9. Design and Licensing Status
Conceptual design is in an advanced stage. The core nucleonic, thermo-hydraulic and heattransfer analyses are being done to optimize the performance and verify the safety analysis.
Nuclear Regulator engagement is planned for 2017 with the aim of commencing the preassessment for licensing in order to reach design certification status at the end of the Concept
Phase.
10. Plant Economics
Since the HTMR-100 is mainly intended to service the needs of isolated geographical areas,
it is not meant to compete economically with large (>1GW) reactors connected to an existing
national grid. The HTR SMR comes into its own right when the various combinations ofelectricity, process heat and water desalination are considered and holds great promise for thelong term economics of users. Preliminary capital and operating cost studies indicate that thisflexibility of the HTMR-100 provides the most efficient energy to isolated communities andindustries. It is well suited to deployment in developing countries with its affordability and
Gen IV safety characteristics.

258

SC-HTGR (AREVA NP, USA)

1. Introduction
The SC-HTGR is a modular, graphite-moderated, helium-cooled, high temperature reactor
with a nominal thermal power of 625 MWth and a nominal electric power capability of 272
MWe. It produces high temperature steam suitable for numerous applications including
industrial process heat and high efficiency electricity generation. The safety profile of the SCHTGR allows it to be collocated with industrial facilities that use high temperature steam.
This can open a major new avenue for nuclear power use. The modular design allows plantsize to be matched to a range of applications. The SC-HTGR concept builds on AREVA’s pastexperience of HTGR projects, as well as on the development and design advances that havetaken place in recent years for modular HTGRs. The overall configuration takes fulladvantage of the work performed on early modular HTGR concepts such as the MHTGR andthe HTR-MODUL.

Figure 1: Nuclear Process Steam Supply
System (Reproduced courtesy of AREVA)

Figure 2: Primary Circuit Layout(Reproduced courtesy of AREVA)

259

2. Target Applications
The SC-HTGR produces high temperature steam suitable for numerous applications
including industrial process heat and high efficiency electricity generation. The HTGR steam
cycle concept is extremely flexible. Since high pressure steam is one of the most versatileheat transport mediums, a single basic reactor module configuration designed to produce hightemperature steam is capable of serving a wide variety of near-term markets. The steam cycle
is also well suited to cogeneration of electricity and process heat.
3. Development Milestones
2009

Project Started

2011

Basic Concept Definition Completed

2012

NGNP Industry Alliance Selected the technology for Commercialization

2012

NGNP Industry Alliance Completed an initial Business Plan for SC-HTGR Commercialization

2014

NGNP Industry Alliance Updated the SC-HTGR Commercialization Plan

2014

Preparation for Pre-Licensing Application Started

2015

Confirmatory design studies in progress (precursor to full Conceptual Design)

4. General Design Description
Design Philosophy
The SC-HTGR is designed around proven helium-cooled and graphite moderated reactor
technology, the heart of which is the TRISO coated fuel particle.
Reactor and Power Conversion Unit
The SC-HTGR reactor is a two-loop modular steam supply system. Each module consists of
one reactor coupled to two steam generators. The steam generators are configured in parallel,each with a dedicated main circulator, as shown in Figure 1. For the reference plant steam
cycle concept, the reactor power level is 625 MW(th).
A steel vessel system houses the entire primary circuit. The reactor vessel contains the reactor
core, reactor internals, and control rods. Each steam generator is housed in a separate steam
generator vessel. Each cross vessel contains a hot duct that channels hot gas from the reactor
outlet to the steam generator inlet. Cool return gas flows in the outer annulus between the hotduct and the vessel wall. The entire vessel system inner surface is bathed in cool reactor inlet
gas, so conventional LWR vessel material (SA-508/533) can be used. This is shown in Figure
2. The reactor vessel is partially insulated while the cross vessels and steam generator vessels
are fully insulated.
Each steam generator is a once-through, helical coil tubular heat exchanger. Feed water enters
the bottom of the steam generator and flows upward inside the tubes, while hot primaryhelium coolant flows downward over the tube bundle. The design of steam generators build
upon the lessons learned at previously operated gas-cooled reactors (e.g., Fort St. Vrain, AVR,
and THTR). Electric motor-powered main helium circulators provide the primary coolant
flow. These variable speed circulators use submerged motors with active magnetic bearings
for simple operation and high reliability.
The reactor inlet and outlet temperatures are 325°C and 750°C, respectively. Thesetemperatures were selected primarily to support the desired steam outlet conditions for thetarget markets. These temperatures also allow the use of standard PWR vessel material, forthe primary vessels without requiring separate cooling or special thermal protection.
260

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

AREVA

Country of Origin

United States of America

Reactor TypePrismatic Block HTGRElectrical Capacity (MW(e))

272

Thermal Capacity (MW(th))

625

Minimum Availability Requirement (%)>90

Design Life (years)

60

Coolant/Moderator

Helium / Graphite

Primary Circulation

2 loops; forced helium circulation – total of 282 kg/s

Steam Generator Power (MW(th))

315 (each)

Main Circulator Power (MW(e))

4 (each)Net Electric Output in All Electric Mode (MW(e))

272Primary System Pressure (MPa)6Main Reactivity Control Mechanism

Control rods (gravity insertion)
Independent reserve shutdown system (gravity insertion)
Large negative temperature coefficient

Reactor Vessel materials

SA508/533 ( typical PWR vessel material)

Coolant Temperature, Core Outlet (°C)

750

Coolant Temperature, Core Inlet (°C)

325

Power Conversion Process

Rankine / Conventional steam cycle

Steam Generator Type

Once-through helical coilSteam Generator Capacity

140.7 kg/s per loop (281.4 kg/s total) – 630 MW(th) total

Steam Conditions

566°C superheated steam at 16.7 MPa

Process Heat / Cogeneration Capability

Yes, High and low temperature or cogeneration

Design Configured for Process Heat Applications

Yes; 100% process heat utilization possible

Passive Safety Features

Passive decay heat removal; Passive shutdown capability

Active Safety Features

Active systems are not required to achieve safe state

Fuel Type/Assembly ArrayUCO TRISO Particle fuel in hexagonal graphite blocks

Fuel Block Height (m)0.8

Number of Fuel assemblies

Annular core, 102 column, 10 blocks / column,

Fuel Enrichment (%)< 20

Fuel Burnup (GWd/ton)< 17

Fuel Cycle (months)½ of the core replaced every 18 months; 21 days outage

Emergency Safety Systems

Passive – No conventional ECCS systems

Residual Heat Removal Systems
Distinguishing Features

Reactor Cavity Cooling System (Passive, water based);
2 natural circulation loops with 2.5MWth heat removal during
accident and 7 days water storage capacity.
Coated particle fuel; passive decay heat removal; passive safety;
high temperature process steam

Modules per Plant

Variable (Reference design is four 625 MW(th) modules)

Design Status

Concept design261

Fuel characteristics
The TRISO coated fuel particle consists of a uranium oxycarbide (UCO) fuel kernel
surrounded by multiple ceramic coating layers that provide the primary fission productretention barrier under all design basis accident conditions. The total fuel inventory includes
roughly 10 billion such particles per core. As shown in Figure 3, 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.

Figure 3: TRISO Fuel Particle, Compact and Block (Reproduced courtesy of AREVA)
Reactor Core Layout
The fuel blocks are configured into a 102 column annular core surrounded by graphitereflector elements as shown in Figure 4. The inner or central reflector also contains graphitereflector elements. Hence the basic core structure is entirely ceramic. This configuration
maximizes the reactor’s passive heat removal capability. The active core is 10 blocks high.

Figure 4: Annular Core Layout (Reproduced courtesy of AREVA)
Cycle Length and Fuel Management
The core cycle length for the 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 U-235
enriched particles with a packing fraction of 0.289 for all fuel elements in the core, and for262

reloads utilizing half-core replacement with fuel blocks having a 15.5 w/o U-235 enrichment
and a packing fraction of 0.279 [1].
Control of local fuel power peaking and limiting of resulting peak fuel temperatures at criticallocations within the fuel block will be accomplished through loading discrete burnable
absorbers, variation of fuel packing fraction, and variation of fuel particle enrichment. This
allows the optimization of the core power distribution in three dimensions and can also be
used to support effective fuel utilization, proliferation resistance and waste reduction.
The high thermal efficiency and high fuel burnup of the SC-HTGR support sustainability forcurrent once-through fuel cycles by minimizing spent fuel volume. The LEU once throughfuel cycle requires about 6.8 MTHM/GWe-yr that equates to a natural uranium feedstock
utilization of about 224 MT/GWe-yr. The SC-HTGR core design is also compatible withvarious more advanced fuel cycles employing fertile/fissile material conversion and recycle
including Th/U, Th/Pu, Pu, and actinide fuel forms.
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. Gravity-driven and diverse reactivity control systems provide further confidenceof the ability to shut down the reactor.
5. Plant Safety Features
Description of safety concept
The primary safety objective of the SC-HTGR design is to limit the dose from accidental
releases so that the US EPA Protective Action Guides are met at an exclusion area boundary
only a few hundred meters from the reactor. To achieve this safety objective, the design uses
the high temperature 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 primary radionuclide retention barrier in the SC-HTGR consists of the three ceramic
coating layers surrounding the fuel kernel that forms a coated fuel particle. The coating
system constitutes a micro-scale pressure vessel around each kernel that has been engineered
to withstand extremely high temperatures without losing its ability to retain fission products
even under accident conditions.
The high temperature capabilities of the massive graphite reactor core structural components
complement the fuel’s high temperature capability. The high heat capacity and low power
density of the core result in very slow and predictable temperature transients even without
cooling. Helium, the reactor coolant and heat transport medium, is chemically inert andneutronically transparent, Helium will not change phase during normal operation or
accidents.
The SC-HTGR is designed to passively remove decay heat from the core regardless ofwhether or not 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 thesystem to 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.
263

Security of the HTGR facility is provided both by the inherent invulnerability of the system
to malicious acts and by the physical optimization of the facility structures to prevent
unintentional access.
Engineered Safety System Configuration and Approach
No powered safety-related systems and no operator actions are required to respond to any of
the accident scenarios that have been postulated for the various modular HTGR concepts
including SC-HTGR throughout their licensing history.
Reactor Cooling Philosophy
The SC-HTGR has three heat removal systems. The two main cooling loops transfer heat tothe secondary circuit during normal operation. When maintenance is being performed on themain cooling loops, a separate shutdown cooling system is available. This system uses a
separate and independent circulator and heat exchanger 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.
If the above two active systems are unavailable, passive heat removal can be used. Heat fromthe core is conducted radially through the graphite reflectors to the core barrel and eventuallyto the reactor vessel. Heat is transferred from the vessel to the Reactor Cavity Cooling
System (RCCS) by thermal radiation and natural convection. This heat removal path remains
effective even if all primary coolant has been lost.

Figure 5: RCCS Flow Path (Reproduced courtesy of AREVA)The RCCS, as shown in Figure 5, is a redundant natural circulation water-cooled system thatmaintains acceptable concrete temperatures in the reactor cavity during normal operation andAnticipated Events, and maintains acceptable fuel, vessel, and concrete temperatures during
Design Basis Accidents. Each independent loop of the safety-related RCCS consists of heat264

collecting panels in the cavity surrounding the reactor vessel connected by a natural
circulation loop to a water storage tank. This loop uses natural circulation for all operating
and accident conditions. A separate, non-safety-related active loop cools the tank during
normal operation. The water in the tank provides the required thermal capacity for aminimum of 7 days of continued cooling during accidents when the active system may not be
available.
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 uses a
vented confinement system. The building provides supplemental fission product retention in
the event of an accident. However, a pressure retaining building such as a light water reactor
containment building is not necessary or technically appropriate due to the excellent fission
product retention performance of the fuel even under extreme accident conditions.
6. Plant Operational Performance
The thermal-hydraulic performance of the SC-HTGR reactor core is linked directly to the
nuclear core design to ensure that the power distribution results in acceptable fueltemperatures for the whole core for all operating modes over the life of the core.
The nominal electricity generation performance of the SC-HTGR system has been evaluated[ 2], taking into account preliminary efficiency estimates for the helium circulators, feedwater
pumps, turbine, generator and other plant electrical loads. The net electrical output from each
625 MW(th) reactor module is 272 MW(e), for a net efficiency of 43.5%. In addition tonominal plant performance, the performance of the SC-HTGR has already been evaluated for
hot arid locations where dry cooling was assumed to be required. Results of this evaluation
indicate that a net electrical generation output of 239 MW(e) is achievable, for acorresponding efficiency of 38.2%.
Steam system equipment can be configured in a variety of ways depending on the specificneeds of the user facility requirements for high temperature process steam, low temperatureprocess steam, and electricity. The secondary and tertiary system of the plant can be easily
customized for each energy user application without affecting the reactor module
configuration or safety case. The steam cycle plant also has excellent load following
characteristics for flexible power operation and grid stabilization for micro-grids with nondispatchable sources. Reactor module power level and steam production can be increased or
decreased on demand and relatively easily. Systems can also shift energy between electricity
generation and heat supply dynamically as real-time load and market conditions vary, allwhile keeping reactor power constant. This provides the maximum utilization of the HTGR
nuclear heat source.
The overall reliability of an industrial energy delivery system using SC-HTGR as a key
component was evaluated for several potential energy plant configurations [3]. The results ofthe study demonstrated that the extreme availability and reliability requirements of a process
heat plant can be met with HTGRs as the primary heat source.

265

7. Plant Arrangement
Each reactor module is located in a separate reactor building. The standard configuration uses
a fully embedded below grade reactor building design as shown in Figure 6. 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 primary functions of the reactor building are to support the NSSS
primary circuit components and to protect the system from external hazards.

Figure 6: Standard Reactor Building Configuration (Reproduced courtesy of AREVA)
Multiple reactor modules can be grouped together on a single plant site. A typical plantlayout might have four reactor modules, although the specific number of modules in an actual
plant, and the timing of construction of each individual module, will depend on the nature of
the application and the customer’s needs. Reactor modules share auxiliary, support andmaintenance systems, but safety systems, including the Reactor Cavity Cooling System andReactor Protection System, are independent.
8. Design and Licensing Status
A concept design and preparatory work for pre-licensing application has been completed.
9. Plant Economics
A business and commercialization plan has been developed as part of the NGNP Industry
Alliance (www.ngnpalliance.org) SC-HTGR commercialization activities.
[1] INEEL/EXT-03-00870, Rev. 1, “NGNP Preliminary Point Design – Results of the Initial
Neutronics and Thermal-Hydraulic Assessments During FY-03”, September 2003.
[2] Lommers, L. J., et. al., “SC-HTGR Performance Impact for Arid Sites”, Proceedings ofthe HTR 2014, Weihai, China, Paper HTR2014-21345, October 2014.
[3] Herd, E., Lommers, L.J., and Southworth, F., “Impact of demand load size on strategies
for reliable process heat supply”, Nuclear Engineering and Design, Vol 251, October
2012.

266

Xe-100 (X-energy, United States of America)
1. Introduction
The Xe-100 is a small-sized pebble bed high temperature gas-cooled reactor with continuous
thermal rating of 100 MW. It features a continuous fuelling regime with low enriched fuelspheres of about 10 wt% entering the top of the reactor going once through the core toachieve a final average burnup of 80000 MWd/tHM after a single passage. The relatively
high burnup causes the bred fissile Pu to be utilized in-situ by about 90%, thus rendering the
spent fuel well depleted. Furthermore, the total used fuel inventory will be stored on-site in adesignated interim storage facility for the life of the plant.

Figure 1: Xe-100 Reactor System Configuration (Reproduced courtesy of X-energy)
2. Target Application
The Xe-100 is intended for electricity production suitable for small or isolated grids. It canalso provide super-heated steam for co-generation, petro-chemical processes, etc. The Xe-100
also provides a scalable platform to increase total power generation from a single site byadding additional reactor modules. Site configurations can consist of 1 to 8 reactor modules
with a small operational staff.

267

3. Development Milestones
2013

Conceptual Design Development

2015

Base Design and Pre-application with USNRC

2017

Application with USNRC

2021

First operation and commercialization

4. General Design Description
Design Philosophy
A major aim of the design is to improve the economics through system simplification,
component modularization, reduction of construction time and high plant availability brought
about by continuous fuelling.
Reactor Core
In reactor design a Cosine-shaped axial flux profile is aimed for. This will yield an optimal
power profile as the spatial neutron reaction rate, RR [cm-3·s-1]:
𝑅𝑅𝑅𝑅(𝑟𝑟⃗) = � ∅(𝑟𝑟⃗, 𝐸𝐸) ∙ 𝜎𝜎(𝐸𝐸) ∙ 𝑁𝑁 ∙ 𝑑𝑑𝑑𝑑
(𝐸𝐸)

with: ∅(𝑟𝑟,��⃗ 𝐸𝐸) ∙ 𝑑𝑑𝑑𝑑 = Integrated neutron flux distribution, together with the reaction rate,
determined by neutron diffusion.
Every time that a fuel sphere reaches the bottom of the core the level of burn-up isdetermined via active measurement of targeted isotopes in a gamma spectrometer through thebuild-up of a preselected long-lived isotope, such as 137Cs.

The OTTO (Once-Through-Then-Out) fuel cycle is the simplest that one can design for, eventhough it delivers a more asymmetric axial flux profile. In this arrangement the burnup
complex can, however, be eliminated as standard equipment, as well as a complicated returnsystem of used fuel from the bottom of the core back to the top. As this implies thereplacement of a complex fuelling/de-fuelling system with a relatively simple one it makes
good sense to consider deploying the system as such.
As noted above the fuel used will be TRISO coated UCO fuel with kernels of slightly smaller
diameter (425 µm) than the usual UO2 fuel (500 µm). The rest of the fuel materials data willremain similar to the German UO2 data. The optimized moderation ratio (NC/NA) has led toa heavy metal loading of around 9 g/pebble. This would enable the Xe-100 under worst casewater ingress scenarios to be able to shut down the reactor with its RCSS.
It is also noted above that a power peaking of 2.4 is observed. Furthermore, the maximum
power rating of the pebbles remains well within the operational envelope of performanceexperimentally determined.
Fuel cycle costs have also been estimated according to the Present Worth method. Subject to
a complex series of economical parameters and assumptions figures around 7.24
Mills/kWh(e) are estimated.

268

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

X-energy LLC

Country of origin

United States of America

Reactor type
Electrical capacity (MW(e))

Modular High Temperature Gas-cooled Reactor(35)

Thermal capacity (MW(th))

100

Expected capacity factor (%)> 95

Design life (years)

40

2Plant footprint (m )100000 (4 reactor modules with one turbine);
160000 (8 reactor modules with two turbines)

Coolant/moderator

Helium / Graphite

Primary circulation

Forced confection; He flow Top  Bottom through core

System pressure (MPa)4oCore inlet/exit temperatures ( C)

260 / 750

Main reactivity control mechanism

Bank 1 (9 Rods) + Bank 2 (9 Rods)

RPV height (m)15

RPV diameter (m)5.15

RPV or module weight (metric ton)–

Configuration of reactor coolant system

Loop type

Power conversion process

Indirect Rankine Cycle

Fuel type/assembly array

Pebbles3

Core volume (m ):27.7

Number of fuel assemblies–

Fuel enrichment (%)

10.61

Fuel burnup (GWd/ton)

79.7

Fuel cycle (months)

On-line refuelling

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (Passive and Active)

Number of safety trains–

Refuelling outage (days)

N/A

Distinguishing features

Online refuelling, core cannot melt and fuel damage
minimized by design, independent fission productbarriers, potential for advanced fuel cycles

Modules per plant

2–8

Target construction duration (months)

24 – 36

Seismic design

0.15g operating limit & 0.3g safe shutdown

Core damage frequency (per reactor-year)

No core melt possible

Design Status

Conceptual Design development269

5. Safety Features
The Xe-100 pebble bed HTR is a simplified design that utilizes well-proven materials andcomponents to expedite the licensing process. It is therefore utilizing a low enriched fuelcycle, with TRISO coatings embedding UCO fuel kernels. Due to its seemingly better
performance parameters this fuel option allows for a dramatic increase in power rating or inburn-up. The decision is pending finalization of the DOE funded UCO fuel qualification
program.
Furthermore, the Xe-100 module is designed for a flexible advanced fuel cycle based oneither uranium or thorium. The double heterogeneous particulate design enables thepossibility to operate under either fuel cycle. Not only does this afford the Xe-100 thepotential entertain a fuel cycle that will produce virtually no reactor grade plutonium andminor actinides, but also yield the potential of utilizing the Xe-100 as Pu-incinerator comingfrom excess weapons’ programs or from a reprocessing process, such as that used by the
French or Japanese.
The intrinsic safety characteristic of the plant is guaranteed by a relatively low power densityof about 3.7 MW/m3, high thermal inertia of the graphitic internals layout and a strong
negative temperature coefficient of reactivity over the total operational regime of the reactor.
This feature is further enhanced via a unique system of independent fission product barriers.
The effects of possible external impacts (such as aircraft or adverse weather) are also
mitigated by the proposed, partially below ground, module layout.

Figure 2: Concept of Independent Fission Product Barriers(Reproduced courtesy of X-energy)
The Xe-100 design is based on the 4 principles of stability depicted below. The fact that the
maximum fuel temperatures will remain below proven safety performance limits of TRISO
fuel means that no core melt will be physically possible and no significant core damage ispossible that would necessitate the evacuation of the general public or cause significant
environmental damage under normal operation and any design basis events.

270

Figure 3: Stability Design: 4 Principles (Reproduced courtesy of X-energy)
6. Plant Arrangement

Figure 4: Conceptual Reactor Module Design (Reproduced courtesy of X-energy)
7. Design and Licensing Status
Conceptual design development and preparation for pre-licensing application.

271

272

FAST NEUTRON SPECTRUM
SMALL MODULAR REACTORS274

LEADIR-PS (Northern Nuclear Industries
Incorporated, Canada)
1. Introduction
LEADIR-PS stands for Lead-cooled Integral Reactor-Passively Safe, a small nuclear reactor
being developed by the Northern Nuclear Industries Incorporated (N2I2) intended to provide
diverse electricity and process heat demands of Canada. LEADIR-PS plants underdevelopment are LEADIR-PS30 (30 MW(th), up to 11 MW(e)), LEADIR-PS100 (100MW(th), up to 39 MW(e), and LEADIR-PS300 (300 MW(th), up to 120 MW(e)). Derivativesof these plants include LEADIR-PS Six-Pack, which features six LEADIR-PS300 modules
that serve a common Turbine Generator, deliveries 720 MW(e), and LEADIR-PS30+ thatincorporates thermal energy storage in the LEADIR-PS30 Balance of Plant to facilitate
electrical outputs of from 5.5 MW(e) to 15 MW(e) over a 24 hour cycle while maintaining
100% reactor power. A simplistic illustration of the LEADIR-PS reactor assembly ispresented in Figure1.

Figure 1: Reactor system configuration of LEADIR-PS (Reproduced courtesy of N2I2)
2. Target Applications
LEADIR-PS100 can serve a wide range of electricity, process heat, and district heating
demands. By providing 12 MPa steam at 370oC, LEADIR-PS units can severe a diverse rangeof process heat, combined heat and electricity, and electricity demands. Energy utilization inprocess heat and combined heat and electricity (CHE) applications can exceed 95% and canbe optimized to suit specific applications. Combined heat and electricity applications canemploy a back pressure turbine that exhausts to a process heat application. As the processapplication serves as the condenser, electrify is generated without any heat rejection to theenvironment. Extraction steam capability can be incorporated into the turbine design tofacilitate the delivery of steam to process users at different pressures. The application
configuration of LEADIR-PS is shown in Figure 2.

275

Figure 2: LEADIR-PS application configuration (Reproduced courtesy of N2I2)
3. General Design Description
Design Philosophy
LEADIR-PS is a molten lead (Pb208) cooled, which in conjunction with the design
configuration precludes a loss of coolant accident (LOCA) and the loss of decay heat removal
capability. Pb208 does not react with graphite, water, or air. LEADIR-PS is graphite
moderated, which via the negative reactivity temperature coefficient, assures inherent reactor
shutdown at elevated temperatures.
Reactor Coolant System
LEADER-PS is a pool type reactor that operates with near atmospheric pressure above thePb208 reactor coolant surface, thereby avoiding high a pressure vessel. It is an integral reactor
with steam generators (SGs) and reactor coolant pumps located within the Pb208 coolant pool
thereby avoiding all high pressure reactor coolant system piping. The primary cooling system
is based on forced circulation during normal operation. The system has natural circulation
capability for emergency condition. The primary reactor cooling system (PRCS) is contained
within the reactor module assembly. The primary reactor cooling system includes two steam
generation modules (SGMs), each consisting of two vertical once through SGs and one
variable speed reactor coolant pump (RCP). Water/steam flows upward within the SG tubes
and the Pb208 coolant flows downward outside of the Steam Generator tubes.
Steam Generator
The SG shell comprises two overlapping concentric cylinders, the SG upper shell and the SG
lower shell that facilitate their relative vertical motion. The resulting ‘floating’ feedwatertubesheet is free to move vertically within the SGM support structure. This configuration
prevents stress in the SG tubes due to thermal expansion and contraction of the SG tubes.
Molten Pb208 has excellent lubrication qualities. The Pb208 coolant flows downward throughthe SGs where the heat of fission is transferred to ordinary water within the SG tubes togenerate superheated steam. On exiting the SGM, the Pb208 coolant enters the suction of areactor coolant pump (RCP). On being discharged from the RCP the Pb208 flows downward.
Nuclear Steam Supply System
The main steam system delivers steam from the SGs to the turbine and to in-plant andexternal process steam users. The main steam system includes the condenser steam discharge
valves, the atmospheric steam discharge valves and the main safety relief valves. In some
applications, two or more LEADIR-PS modules may be configured to supply steam to asingle turbine-generator. In this case, dedicated feedwater control is provided to each
LEADIR-PS unit.
276

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

Northern Nuclear Industries Incorporated

Country of origin

CanadaReactor type

Liquid metal cooled fast reactor

Electrical capacity (MW(e))

39

Thermal capacity (MW(th))

100

Expected Availability Factor (%)–

Design Life (years)

60

2Plant Footprint (m )–Coolant/moderator

Pb208 / Graphite

Primary circulation

Forced circulation

Number of reactor coolant pump

2 (Vertical Centrifugal)

System pressure (MPa)12oCore inlet/exit temperatures ( C)

360 / 560

Main Reactivity Control Mechanism

Control rods, on-power fuelling system

RPV Height (m)10.2

RPV Diameter (m)3.2

Module Weight (metric ton)–

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine Cycle

Fuel Type/Assembly Array

TRISO coated particle

Fuel material

U235 oxide

Fuel pebble diameter (mm)60

Number of Pebble Bed cells

3

Number of fuel pebbles in core

94,500Fuel Enrichment (%)

9.8

Fuel Burnup (GWd/ton)

86

Fuel Cycle (months)–

Breeding ratio–

Cogeneration Capability

Yes

Approach to Engineered Safety Systems

Passive

Number of safety trains–

Refuelling Outage (days)–

Distinguishing features

Eliminates the need of pressure vessels; low boiling point coolants;
Multi-purpose reactor for electricity, process heat, district heating
and combined heat and electricity.

Modules per plant

1

Target Construction Duration (months)

24

Seismic design–

Predicted core damage frequency (per reactor year)–

Design Status

Conceptual design277

Reactor CoreLEADIR-PS utilizes a core design based on the Pebble Bed High Temperature Gas Reactor(HTGR), which was developed in Germany in the 1960s for the AVR-15 and THTR300, andwhich has been adopted by China for their HTGR. LEADIR-PS subdivides the core into three(3) pebble bed cells, as shown in Figure 3. Each pebble bed cell, essentially a small reactor
consisting of a cylindrical vertical cavity in the graphite reactor structure occupied by Fuel
Pebbles, has an output of 34 MW(th). The number of pebble bed cells depends on the reactor
output (one in LEADIR-PS30, three in LEADIR-PS100, and nine in LEADIR-S300). The
fuel pebble contains 16,500 TRISO particles. There are 32,500 Fuel Pebbles in a pebble bed
cell. The graphite moderator, reflector and core barrel constitute the reactor core assembly(RCA). The reactor core assembly is housed within the Pb208 filled primary reactor vessel(PRV). Pb208 has a relatively low thermal neutron capture cross-section which contributes to
excellent fuel utilization. The fuel pebbles and graphite moderator/reflector blocks arebuoyant in the Pb208 coolant.

Figure 3: Reactor core plan section (Reproduced courtesy of N2I2)LEADIR-PS utilizes 0.92 mm diameter TRISO fuel particles (0.6 mm diameter uraniumoxide kernel with multiple coatings), with an enrichment of 9.8% to obtain sufficient uranium
mass in the reactor core and to provide for high burnup. New fuel pebbles contain TRISO
particles with a small amount of burnable neutron absorber that to serves to limit the heatgenerated in the pebble during their initial period in the reactor core. The coatings retain
fission products under all potential conditions thereby precluding a radioactive release to theenvironment. The interlocking graphite moderator and reflector blocks forms the principal
structure of the reactor core and define the cylindrical pebble bed cells. The graphite block
structure is designed to accommodate thermal expansion and to minimize the creation ofpassageways resulting from graphite block distortion caused by radiation. The graphite
structure is a vertical cylinder and together with the reactor barrel (RB) constitutes the RCA.
Reactivity Control
The three pebble bed cells are neutronically closely coupled. The small LEADIR-PS100 coreis stable in all modes. Reactor control takes advantage of the strong negative reactivity
temperature coefficient of the graphite reflector/moderator and the high heat capacity of thecore. The on-power fuelling system is the primary mechanism for reactor reactivity control.
Operational reactor control is provided by the reactor control system (RCS). The RCS
monitors and controls all functions that are related to reactivity control, including the onpower refuelling system, except for the reactor shutdown system. The fully automatic RCS
utilizes dual redundant computers that incorporate numerous checks and safeguards.
278

Reactor Pressure Vessel and Internals
The primary reactor vessel (PRV) is housed within the secondary reactor vessel (SRV). The
annular space between the PRV and the SRV, called the annulus, is filled with lead enrichedin neutron absorbing lead isotopes. This lead consists of the ‘tails’ from the Pb208 isotope
separation process plus ordinary lead. A vertical cylindrical divider within the annulus, with
openings at the top and bottom, divides the annulus into two annular cylindrical sections. The
integral pool reactor configuration in combination with the high boiling point Pb208 coolant
avoids the need for a pressure vessel and high pressure primary RCS piping, thereby
eliminating the potential for high energy pressure vessel or piping rupture.

Figure 4: Simplified LEADIR-PS100 Reactor Module Assembly Elevation Section(Reproduced courtesy of N2I2)
4. Safety Features
LEADIR-PS provides an unprecedented level of safety, which includes:
•
•
•
•
•
•
•

No potential for a reactivity transient without shutdown/scram;
No potential for a LOCA, loss of heat sink and containment failure;
No potential for pressure vessel or high pressure RCS failure;
No potential for hydrogen generation, by coolant disassociation or zirconium water
reaction;
No potential safety impact by incompetent or malicious operator/maintenance staff or
third party via the control system or component degradation;
No requirement for operator action, valve repositioning or any power source;
No requirement for any above ground structures or components to be functional.

LEADIR-PS is located in a below grade silo for protection from external events and toprovide an assured long term heat sink. The reactor is inherently and passively safe requiring
no power sources, no operating systems, not repositioning of valves and no operator actions
to assure safety. LEADIR-PS remains inherently safe in the event that all above grade
systems and structures are lost, and in the unlikely event of a fire above the reactor assembly.

279

Engineered Safety System Approach and Configuration
The passive shutdown system incorporates shutdown rods (SDRs) that drop by gravity withspring assist when power is removed from electromechanical clutches. SDRs deployment by
the reactor shutdown system is backed up by passive initiation by the passive shutdown
initiation system (PSIS) that removes power from the clutches when a specific core outlet
temperature is reached. Inherent shutdown capability is provided by the negative temperature
coefficient of the graphite, thereby assuring reactor shutdown in the event that both thereactor shutdown system and PSIS fail when needed. The high structural capability ofgraphite at high temperature (maximum strength at 2500oC) assures core integrity under all
postulated conditions. LEADIR-PS safety, including inherent shutdown and passive decay
heat removal capability is maintained even if the above ground portion of the Reactor
Protective Structure is extensively damaged or demolished.
Decay Heat Removal System
After the reactor is shutdown, when the normal decay heat removal mechanism utilizing thesecondary system is not operable by any reason, core submergence and natural circulation of
the molten lead coolant is assured. As lead in the annulus is solid and would remain in place
during normal reactor operation even if the secondary reactor vessel is not present thissituation is extremely unlikely. Passive decay heat removal is assured for all credible events
for an indefinite period of time without the need to reposition valves or other devices andwithout any power or any operator actions. Although provided, natural circulation of molten
lead coolant is not required to assure decay heat removal.
Containment System
The coatings retain fission products under all potential conditions thereby precluding aradioactive release to the environment. No reactor containment is required. The location of
the reactor assembly below grade and submerged in lead in combination with the passive
decay heat removal capabilities makes the core highly resistant to external events. Additional
protection is afforded by the above ground reactor protective structure.
5. Plant safety and Operational Performances
Heat removal is via the SGs utilizing the main feedwater system (MFWS) and the maincondenser (MC) during most periods. Turbine steam bypass to the condensers enables
continued reactor operation at 100% full power for up to 5 minutes and indefinite operation at
up to 60% powers in the event that the Turbine-Generator is unavailable. LEADIR-PS100
operating characteristics can be summarized as:
•
•
•
•
•
•
•

280

Black start capability independent of the electricity grid;
100% power operation without an electricity grid connection for pure process heatapplications and operation at up to 100% power as the only electricity supply to theelectricity grid;
The reactor can maintain sustained operation at any power level between 10% and100% of full power;
Reactor control is normally in the reactor following load (electricity and/or processheat) mode;
The power design increase rate for the reactor and reactor systems is 0.2 percent offull power per second between 10% and 100% power;
The power reduction design rate for the reactor systems is 0.2 percent of full powerfrom 100% power to 60% power and 0.05 percent per second below 60% power;
More rapid load reductions (process heat or electricity) results in steam bypass to thecondenser until reactor power drops to the required level.

6. Plant Arrangement
The compact layout of a single module LEADIR-PS100 is illustrated in Figure 5. The
principal structures are the reactor protective structure (RPS), the turbine building (TB), theelectrical and control building (E&C), the radioactive storage building (RSB), the auxiliary
services building (ASB), and the access corridor. The E&C building structure is integrated
with that of the TB and is separated from the TB by a waterproof and fireproof wall. The
E&C building is seismically and tornado qualified to in accordance with site conditions.
Additional onsite facilities include the switch yard, and the pump house if water cooled Main
Condensers is used. An accommodations building (AB) which provides sleeping, kitchen,lunch, and relaxation facilities for the maintenance crews when on site is provided for
LEADIR-PS100 units at remote or isolated locations. The layout of plants with multiple
LEADIR-PS modules is similar to that of the single module plant, but employs an integrated
RPS design, and may, depending on project specific energy demands, have a common
turbine-generator serving more than one LEADIR-PS100 module.

Figure 5: LEADIR-PS general plant arrangement with a single unit
(Reproduced courtesy of N2I2)
Reactor Building
The reactor building and fuel storage area are equipped with a full monitoring system withclosed circuit monitoring system (CCTV). Figure 5 shows the overview of the reactor
building, auxiliary building, and compound building arrangement of LEADIR-PS.
Balance of Plant:
Turbine Generator Building
The TB is a three level structure located to the ‘A’ side of the E&C building. The bottom
level of the TB, located below grade, houses the plant water systems, the services boiler(SB), and the main condenser (MC) if a water cooled main condenser is used. The
second level of the TB accommodates the turbine-generator (TG), the TG support
systems including the lube oil system, the generator hydrogen system, the diesel
generators and mechanical shop area.
281

Electric Power Systems
These systems include the main generator, main transformer, unit auxiliary transformers,
stand-by auxiliary transformers, diesel generators, and batteries. The electrical systems,
including the class 1E and non-class 1E, are based on a two train approach. The electricalservices (ES) provide power for lighting, heating, and the operation of tools andmachinery to all plant buildings as required. Power to ESs is provided by the Group A
and Group B power supplies. The ESs is connected to their power supply only when amaintenance team is present. Power supply to the ESs can be established remotely by thecentral operations facility.
7. Plant Economics
The advanced delivery technologies of LEADIR-PS100 facilitate short and secure
construction schedules. For LEADIR barge delivery subsequent to the first unit, theconstruction schedule is 24 months. This includes 15 months for the fabrication of the
LEADIR barge, 1 month for the delivery of the LEADIR barge and 8 months for thecompletion of site work and commissioning. Site preparation and the construction of site
support facilities proceeds in parallel with LEADIR barge fabrication. The schedule for fully
modularized on-site construction, for LEADIR-PS units subsequent, to the first unit, is 30
months.

282

4S (Toshiba Corporation, Japan)
1. Introduction
The 4S (super-safe, small and simple) is a small sodium-cooled pool-type fast reactor withmetal fuel. Being developed as distributed energy source for multi-purpose applications, the4S offers two outputs of 30 MW(th) or 10 MW(e) and 135 MW(th) or 50 MW(e),respectively. These energy outputs are selected from the demand analyses. The 4S is not abreeder reactor since the blanket fuel, usually consisting of depleted uranium located aroundthe core to absorb leakage neutrons from the core to achieve breeding of fissile materials, is
not present in its basic design. 4S reactor cores are designed to have a lifetime of 30 years forthe 30 MW(th) core and 10 years for the 135 MW(th) core. The movable reflector
surrounding the core gradually moves, compensating for the burnup reactivity loss over thecore lifetime. The plant electric power can be controlled by the water–steam system, whichmakes the reactor applicable for a load follow operation mode.

Figure 1: Reactor system configuration (Reproduced courtesy of Toshiba Corporation)
2. Target Application
The 4S is designed for electricity supply to remote areas, mining sites as well as for nonelectric applications. The plant can be configured to deliver hydrogen and oxygen using theprocess of high temperature electrolysis. This process can be performed without producing
environmentally disadvantageous by-products, such as carbon dioxide. Two kinds of systems
for non-electric applications can be incorporated in the 4S;••

Seawater desalination system: the 50MW(e) 4S plant can produce fresh water at a rateof 168000 m3/day;Hydrogen and oxygen production system: hydrogen production rate for the 10MW(e)and 50 MW(e) 4S is 3000 Nm3/h and 15000 Nm3/h respectively.

Combinations of these systems and the turbine generator system as balance of plant (BOP),
283

including the capacity of each system, would be determined to meet the actual needs at any
particular site.
3. Development Milestones
2007

Licensing activity for the 4S design initiated with the U.S. Nuclear Regulatory Commission(U.S.NRC)

2008

Completion of 4 times public meetings as pre-application review with U.S.NRC

2013

Completion of submitting 14 technical reports to U.S.NRC

Air Cooler

Turbine
EMF

Generator

Separetor

SG
RVACS IHX

EMP

Condenser

EMP

Pump
Core

Heater

FWP

HeaterFigure 2: Simplified schematic diagram of 10 MW(e) 4S plant(Reproduced courtesy of Toshiba Corporation)
4. General Design Description
Design Philosophy
The 4S reactor is an integral pool type with all the primary components installed inside thereactor vessel (RV). Major primary components consist of intermediate heat exchangers(IHX), primary electromagnetic pumps (EM pump), moveable reflectors which form a
primary reactivity control system, the ultimate shutdown rod, radial shielding assemblies, thecore support plate, coolant inlet modules and fuel subassemblies. The 4S design is optimized
to achieve the following design targets based on conventional sodium-cooled fast reactor
technology with some innovative technologies:
•
•
•
•
•

284

Improvement of the public acceptance and safety: all reactivity feedback are zero or
negative and inherent safety features with metal fuel;Minimization of fuel cost and operation and maintenance (O&M) cost; ensuring
enhanced proliferation resistance; long refuelling intervals (30 years for the 10MW(e)-4S and 10 years for the 50 MW(e)-4S);
Taking into account political circumstances: use of uranium fuel with the enrichment
by U235 less than 20% (by weight);
Securing fuel integrity under long-life operation of the core: adequate fuel burn-up;
Minimization of construction costs: reduction of core size.

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Toshiba Corporation

Country of origin

Japan

Reactor type

Liquid metal cooled fast reactor (Pool type)

Electrical capacity (MW(e))

10Thermal capacity (MW(th))

30

Expected capacity factor (%)> 95

Design life (years)

60

2Plant footprint (m )10000

Coolant/moderator

SodiumPrimary circulation

Forced circulation

System pressure (MPa)Non-pressurizedoCore inlet/exit temperatures ( C)

355 / 510

Main reactivity control mechanism

Axially movable reflectors / Fixed absorber

RPV height (m)24

RPV diameter (m)3.5

RPV (metric ton)

100

Configuration of reactor coolant system

Integral type (pool type)

Power conversion process

Indirect Rankine cycle

Fuel type/assembly array

Metal fuel (U-Zr alloy) based on enriched uranium

Fuel assembly active length (m)2.5

Number of fuel assemblies

18

Fuel enrichment (%)< 20

Fuel burnup (GWd/ton)

34

Fuel cycle (years)

N/A

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (Active + Passive) system

Number of safety trains

3

Refuelling outage (days)

N/A

Distinguishing features

Modules per plantCore lifetime of ~30 years without on-site refuelling, power
control by the water/steam system without affecting the core
operation, passive walkaway safety.
1

Target construction duration (months)

24 months for on- site construction period

Seismic design

Seismic isolator

Core damage frequency (per reactor-year)

1.7E–8Design Status

Detailed design

Nuclear Steam Supply System
The nuclear steam supply system (NSSS) consists of the primary cooling system, theintermediate heat transport system and the water/steam system. The intermediate heattransport system has an EM pump, piping and a steam generator (SG). The SG is a helicalcoil type with wire-meshed double-wall tube to prevent a sodium-water reaction in the eventof the tube failure.

285

Figure 3: Under grade reactor building of the 4S plant(Reproduced courtesy of Toshiba Corporation)
Reactor Core
The core and fuel are designed to eliminate the need for refuelling during approximately
thirty (30) years for the 10 MW(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 during thirty (30) years by axially moving reflectors installed at the outside of thecore, upward from the bottom. No reloading or shuffling of fuel is required during the wholecore lifetime. The fuel element (fuel pin) consists of fuel slugs of U-Zr alloy, bonding
sodium, cladding tube, and plugs at both ends. A gas plenum of an adequate length is locatedat the upper region of the fuel slugs. In the fuel subassembly, fuel pins are assembled and a
top shield is installed to prevent activation of the EM pumps and the secondary sodium in the
IHX. 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 RVACS.
Reactivity Control
The reactivity control during normal operation is by the axial movement of reflectors andusing fixed absorbers. The movable reflector surrounding the core gradually moves,compensating the burn-up reactivity loss over the 30 years lifetime. Therefore, the reactivity
control is unnecessary at a reactor core side and this is an important factor to simplify thereactor operation. The transient overpower is prevented by the limitation of high-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, theprimary EM pumps, the moveable reflectors which form a primary reactivity control system,
the ultimate shutdown rod which is a back-up shutdown system, radial shielding assemblies,
core support plate, coolant inlet modules and fuel subassemblies. The RV provides a primaryboundary for the primary sodium coolant, and is designed with a pressure/temperature of 0.3
MPa/550oC. The design lifetime of the RV is 60 years as well as the other components.
Reactor Coolant System
The primary sodium circulates from the EM pumps downward, driven by its pump pressure,and flows through radial shielding assemblies located in the region between the RV and the286cylindrical dividing wall. The coolant flow changes its direction at the bottom of the RV andthen goes upward, mainly into the fuel subassemblies and partly into the movable reflectors.
The coolant flow is distributed appropriately to fuel subassemblies of each 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 are also cooled so that the temperature becomes sufficiently low and thetemperature distribution is flattened to maintain integrity through the plant life time. The
coolant gathers at the hot plenum after flowing through the fuel subassemblies and thereflectors. The heated primary sodium then goes into the IHX to transfer heat to thesecondary 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 the secondary 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 circulatedby 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 tothe steam turbine generator.
Steam Generator
The 4S adopts a once through type double-wall tube SG with failure detection systems. The
heat transfer tube of the SG is a double-wall type. Between the inner and outer tube, wire
meshes are installed, which are filled with helium, to detect one side tube failure prior to
failure of the other side tube. It enables to prevent sodium-water reaction.
Pressurizer
The 4S is a sodium cooled fast reactor that does not need to pressurize inside the primarycoolant boundary. Hence it has no in-vessel pressurizer.
5. Safety Features
The philosophy of the 4S safety concepts is to put an emphasis on simplicity achieved using
passive and inherent safe features as a major part of the defence in depth (DiD) strategy. The
ultimate objective of the 4S safety concept is to eliminate the requirement of evacuation as an
emergency response measure. The 4S safety concept provides for three functions in eachphase of the abnormal operation or an accident: prevention; mitigation; and confinement ofradioactive material. The safety systems of the 4S consist of: redundant shutdown system;
passive decay heat removal system without external power supply; emergency power system;
and a reinforced reactor building. The active and passive/inherent safe features of the 4S are
applied with the following main objectives:
•
•
•
•
•

Reduce the probability of component failure;
Prevent core damage in accidents;
Two fully passive heat removal systems;
Confine the radioactive materials;
Prevent sodium leakage and to mitigate the associated impact if it occurs.

Engineered Safety System Approach and Configuration
In addition to the inherent safe features, there are two independent systems for reactor
shutdown. The primary shutdown system provides for a drop of several sectors of the287

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. Either the reflector or shutdown rod is capable of enough negative
reactivity to shutdown the reactor.

Figure 4: Engineered safety system schematic for decay heat removal of 4S(Reproduced courtesy of Toshiba Corporation)
Decay Heat Removal System
The water/steam system is available for normal shutdown heat removal. The decay heat of thecore is transferred to water/steam system via the intermediate heat transport system by forced
convection and is finally removed from condenser. For decay heat removal during water
steam system is not available upon accidents, two independent passive systems are provided;
the reactor vessel auxiliary cooling system (RVACS) and the intermediate reactor auxiliary
cooling system (IRACS). The RVACS is completely passive and removes decay heat from thesurfaces of the guard vessel (GV) using natural circulation of air. There is no valve, vane, or
damper in the flow path of the air; therefore, the RVACS is always in operation, even when
the reactor operates at rated power. Two stacks are provided to obtain a sufficient draft. The
IRACS removes decay heat by air cooler which is arranged in series with the secondary
sodium loop. Heat is removed by forced sodium and air circulation at the IRACS whenelectric power is available. In addition, the IRACS can also remove the required amount ofheat solely through natural circulation of both air and sodium during loss of power events.
Emergency Core Cooling System
Pool-type sodium cooled fast reactors usually have RV and GV to keep core immersed inprimary coolant even if RV failure occurs since GV keeps coolant in it. Also, primary coolant
cannot be forced out from RV since there is no high pressure in RV unlike light water reactors(LWR) and all primary coolant is contained in RV. In addition, 4S has passive reactor coolingsystem, RVACS as mentioned above. It plays a role as an emergency core cooling system.
There is no need to have LWR-like emergency core cooling system.
Containment System
The 4S adopts a cylindrical/spherical containment system. The containment system consistsof the GV and the top dome, which covers the upper region of the RV, a shielding plug andthe 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 isprovided inside the top dome.

288

6. Plant Safety and Operational Performances
The 4S is designed to operate safely without active involvement of the plant operators. The
design features to support such operation are:••••Burn-up reactivity swing is automatically compensated by the fine motion reflectors;
There is no need in reloading and shuffling of fuel in the course of 30 years for the 10MW(e)-4S;
A reduction in maintenance requirements achieved by adopting static devices such as
EM pumps or static devices continuously monitored by simple systems;
Reduction of in-service inspections achieved by taking advantage of the nonpressurized 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 Systems
The instrument and control system consists of safety related and non-safety related system.
The system related to safety includes the reactor protection system (RPS), the engineeringsafety feature actuation system (ESFAS) and the remote shutdown system (RSS). Thesesystems have the safety class1E instruments. The RPS is plant protection system to initiate
reactor trip at abnormal plant operation condition. The trip parameters are as follows:
•
•
•
•
•
•

Reactor core neutron flux;
Liquid sodium level in the RV;
Supply voltage/current of main/intermediate loop EM pumps;
Primary outlet temperature of IHX;
Voltage of power line;
Seismic acceleration.

The RSS provides means to shut down the reactor from outside the main control room. The
system related to non-safety consists of the plant control system (PCS) and the interface
reactor system (IRS).
8. Plant Arrangement
The plant layout of the 4S is optimized to meet various functional needs; the requirements forsafety, radiation zoning, piping and cabling; construction requirements; access and security
considerations. The general philosophy of the 4S plant layout is as follows:
•
•
•
•
•

Efficient space utilization and minimization of volume of the buildings;
Horizontal seismic isolation for the reactor building;An embedded reactor building, securing that the reactor is earth-sheltered;Lightweight buildings to assure a high degree of transportability in construction;
The secondary sodium loop area is categorized as a non-radiation-controlled area; toachieve this, a sufficient shielding for the IHX is provided.

Reactor Building
The 4S is a land-based nuclear power station with the reactor building embedded
underground for security reasons, to minimize unauthorized access and enhance inherent
protection against extreme external events. The reactor building is supported by horizontal
seismic isolators, reinforced and protected from massive water invasion by keeping its watertightness. The reactor building including the concrete silo can be used for more than 60 years.

289

Dry cask
storage area

Figure 5: The 4S plant layout diagram (Reproduced courtesy of Toshiba Corporation)
Control Building
The control room is settled in the reactor building which is supported by the horizontal
seismic isolators and embedded underground.
Balance of Plant:
The BOP including a steam turbine system is located at ground level.
Turbine Generator Building
The 4S plant consists of one reactor and one turbine generator system. Superheatedsteam is supplied from the steam generator to the turbine.
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 gridis also connected to the unit auxiliary transformer (UAT). The PMG supplies the power
to the onsite power subsystem via the UAT. The two class 1E buses are separated fromeach 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. Design and Licensing Status
Licensing activities for the 4S design initiated with the U.S.NRC in 2007. In pre-application
review, four (4) meetings had been held in the past and fourteen (14) technical reports havebeen submitted to NRC. Toshiba is conducting the detailed design and safety analysis fordesign approval. In parallel, Toshiba continues to look for customers.
10. Plant Economics
To be a viable option for power generation in remote areas, the 4S should provide acompetitive cost of electric power. The target plant construction cost is US$ 10000/kWe andthe O&M cost of 4.5¢/kWh. The target construction cost could be achieved by the simplified
and standardized design, shop fabrication, modular construction, short construction time, andmass production. The O&M cost could be achieved by long-refuelling interval, and lowmaintenance requirements of components such as EM pump.

290

BREST-OD-300 (NIKIET, Russian Federation)
1. Introduction
BREST-OD-300 is a lead cooled fast reactor fuelled with uranium plutonium mononitride(PuN–UN) that uses a two circuit heat transport system to deliver heat to a supercritical steam
turbine and generate electricity of 300 MW for experimental and demonstration purposes.
BREST-OD-300 is the reference Russian design of a medium-size lead cooled fast reactor.
The lead-cooled fast reactor is one of the alternative fast reactors under development in the
country. Russian Federation has the operational experience on the use of lead-bismuth
eutectic alloy in power reactors for submarine propulsion, using the SVBR-80 technology.
This experience is being incorporated in the development of lead cooled fast reactors. The
present goal of the project is to implement all necessary R&D in order to finalize the detailed
design of the BREST˗OD-300.

Figure 1: Reactor system configuration of BREST-OD-300(Reproduced courtesy of NIKIET)
2. Target Applications
The BREST-OD-300 power unit is designed as a pilot and demonstration unit intended forstudying the reactor facility operation in different modes and optimizing all processes andsystems that support the reactor operation. After operational tests, the unit will becommissioned for electricity supply to the grid. The multi-purpose application configurationof BREST is shown in Figure 2.

291

Figure 2: BREST multi-purpose application configuration (Reproduced courtesy of NIKIET)
3. Development Milestones
1995

Conceptual design development initiated2002

A feasibility study of the BREST-OD-300 NPP with an On-Site Nuclear Fuel Cycle(OSNFC) for the Beloyarsk NPP Site
First of a kind engineering demonstration plant2020s

4. General Design Description
Design Philosophy
BREST-OD-300 is a pool type reactor design not requiring metal vessel, hence, not limiting
reactor power. The objective of the design is to eliminate severe accidents; complete fuelbreeding (equilibrium mode) for self-sustaining; integral-type arrangement of the first circuit
to avoid output of coolant outside the reactor vessel, to eliminate loss of coolant; use of lowactivated coolant with high enough boiling temperature, without rough interaction with waterand air in the case of depressurizing of the circuit. The reactor facility has a two-circuit steam
generating power unit that includes a reactor, steam generators (SG), main coolant pumps(MCP), fuel assembly (FA) loading system, control and protection system (CPS), concrete
vault with a heat shield, steam-turbine unit, cool-down heat removal system, reactor heat-up
system, reactor overpressure protection system, gas purification system and other auxiliary
systems.
Reactor Coolant System
BREST-OD-300 uses a mixed integral primary circuit configuration as the SGs and MCPs
are installed outside the reactor central vessel. As compared to the integral traditional fast
reactor designs, the BREST concept has reduced dimensions and volume of the primarycircuit. The reactor and the SGs are located in the thermally shielded concrete vault, without
using a metal vessel. The concrete temperature is kept below the allowable limit by means of
natural air circulation. 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 level292

in the suction chamber and supplied to the free level of the annular pressure chamber. The
lead further goes down to the core support grid, flows upward through the core where it isheated up to the temperature of 540°C, and is supplied to the common “hot” coolant drainchamber. Then it goes up and flows over the SG inlet cavities and inter-tube space via the
distribution header nozzles. As it goes down in the inter-tube space, the LC transfers heat tothe 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 from it is pumped again to thepressure chamber shown in Figure 3.

Figure 3: Schematic Flow Diagram of BREST-OD-300 (Reproduced courtesy of NIKIET)
The absence of high pressure in the primary lead circuit and a relatively high lead freezing
temperature contribute to crack self-healing, which excludes loss-of-core-cooling accidents
and outflow of radioactive lead from the reactor vessel. Lead circulation through reactor coreand steam generator takes place due to the difference between the levels of cold and hot
coolant developed by the pumps, rather than the head developed by the pumps. Nonuniformity of lead flow through the steam generators with one of few pumps shut down is
excluded, in so doing flow inertia in fast pump shutdown is provided by equalizing coolant
levels in discharge and suction chambers. As for primary coolant flow system, while flowing
the coolant reaches free level twice, that results in surfacing and escaping majority of leadvapour bubbles, being generated in accidents caused by steam generator tube
depressurization.
Nuclear Steam Supply System
The primary cooling system is based on forced circulation during normal operation. The
system has natural circulation capability for emergency condition. The integral BREST
design of high-boiling point (~2000 K) and low-activated lead coolant, does not require highpressure in the primary circuit, and excludes the potential loss of coolant accidents.

293

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology Developer

NIKIET

Country of origin

Russian Federation

Reactor type

Liquid metal cooled fast reactor

Electrical capacity (MW(e))

300

Thermal capacity (MW(th))

700

Steam production rate (t/hr)

1480

Expected Availability Factor (%)

82

Design Life (years)

30

2Plant Footprint (m )65 x 74

Coolant/moderator

Lead

Primary circulation

Forced circulation

Number of reactor coolant pump4

System pressure (MPa)Low pressure operationoCore inlet/exit temperatures ( C)420/540

Main Reactivity Control Mechanism

Shim and automatic control rods (∆ρ ≈ 14 βeff)

RPV Height (m)14.14RPV Diameter (m)6.88

Module Weight (metric ton)

8600

Configuration of Reactor Coolant System

Integrated

Power Conversion Process

Indirect Rankine Cycle

Fuel Type/Assembly Array–

Fuel material

PuN-UN

Fuel Assembly Active Length (m)1.1

Number of Fuel Assemblies

145

Fuel Enrichment (%)~ 13.5

Fuel Burnup (GWd/ton)

61.45

Fuel Cycle (months)

60-72, with partial refuelling – 1/year

Breeding ratio

1.05

Cogeneration Capability

Yes

Approach to Engineered Safety Systems••

Passive and time-unlimited direct residual heat removal fromthe lead circuit system via natural air circulation with heat
discharge into the atmosphere
Emergency protection rods (∆ρ ≈ 6 βeff)

Number of safety trains–

Refuelling Outage (days)–

Distinguishing features

High level of inherent safety due to natural properties of the lead,
fuel, core and cooling design

Modules per plant

1

Target Construction Duration (months)–

Seismic design

VII-MSK 64

Predicted core damage frequency (per reactor year)–

Design Status

Detailed design; for potential start-up in early 2020s

294

Reactor Core
The reactor of a pool-type design has an integral lead circuit accommodated in one central
and 4 peripheral cavities of the concrete steel-lined 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-MCP units, heat exchangers of theemergency and normal cool-down systems, filters and auxiliary components. The cavities
have hydraulic interconnection. The BREST reactor core has a high-boiling point of~1725oC, radiation-resistant, low-activated lead coolant, which is chemically inert with waterand air, does not require high pressure in the primary circuit, and excludes the potentiality ofaccidents with a loss of coolant and heat removal, fires and explosions in a contact with theenvironment. The lead coolant properties in combination with a dense, heat-conducting
nitride fuel provide conditions for complete plutonium breeding in the core (CBR ≥ 1). That
results in a small operating reactivity margin (Δρ < βeff) and enables power operation
without prompt-neutron reactor power excursions.

Figure 4: The BREST-OD-300 reactor core (Reproduced courtesy of NIKIET)
The adopted fuel is mixed mononitride (UN-PuN) which exhibits high density (14.3 g/cm3)and high conductivity (20 W/m·K) and is compatible with lead and the fuel cladding ofchromium ferritic-martensitic steel. In order to reduce the fuel temperature and thus make therelease of fission products from the fuel relatively low, the gap between the fuel and the
cladding is filled with lead which ensures a suitable thermal fuel-coolant contact. To provide
a significant coolant flow area, increase the level of power removed by natural lead
circulation, reduce the coolant preheating temperature and, primarily, exclude the coolinglosses in the damaged FA in case of local flow rate blockage, all core FAs do not haveshrouds. The FA design allows radial coolant overflow in the core which prevents
overheating of the damaged FA. The FA design allows radial coolant overflow in the core thatprevents overheating of the damaged FA. Better radial uniformity of FA power levels andcoolant heating is ensured by regionalized fuelling and lead flow rates by placing assemblies
with smaller-diameter rods in central regions, and larger-diameter in peripheral regions. The
adopted fuel exhibits high density and high heat conductivity, and is compatible with lead andthe fuel cladding of chromium ferritic-martensitic steel. The reflector assemblies have leakproof cans. The first row of these assemblies is used as control element channels and theassemblies in rows 2 to 4 can contain long-lived iodine and technetium for transmutation and295

also contain Sr90 and Cs137 as a stable heat source for prevention of hypothetical coolant
freezing. Instead of traditional flattening of the radial power density distribution by means ofdifferent fuel enrichments, this is achieved by means of a core composed of three zones with
fuel elements having different diameters, but the same plutonium content. Such a method
guarantees a suitable flattening of the lead temperatures at the core outlet, as well as therespect of the maximum fuel cladding temperatures. To reduce the neutron leakage, to betterlevelling the neutron flux, ensure the operation conditions without cycle reactivity change
and attain the complete fuel breeding in the core (CBR ~ 1), traditional uranium shields havebeen replaced with an effective lead reflector whose albedo characteristics are better thanthose of uranium dioxide.
Reactivity Control
Reactivity control during normal operation is achieved using shim and automatic control rods(∆ρ ≈ 14 βeff). Emergency protection rods are also provided (∆ρ ≈ 6 βeff).
Secondary Circuit
The use of chemically inert, high-boiling molten lead in the primary circuit allows adoptionof a two-circuit unit configuration, with a supercritical steam system as secondary circuit.
The secondary circuit is a non-radioactive circuit consisting of SGs, main steam lines, afeedwater system and one turbine unit with supercritical steam parameters. A standard K-300240-3 turbine unit with three-cylinder (HPC+MPC+LPC) steam condensation turbine withintermediate steam superheating and a rotation speed of 3000 rev/min is used. The nominal
steam flowrate to the turbine is about 1000 t/h. Oxygen neutral water at supercritical pressureis used in the secondary loop. Such provision does not require removal of oxygen and makethe deaerator in the secondary circuit unnecessary. The intermediate steam superheating
system contains two steam-to-steam heat exchangers (SSHX). The SSHX heating fluid issteam extracted from the main steam lines upstream of the turbine at a flowrate of 600 t/h.
Downstream of the SSHX, all cooled fluid flowrate is supplied via throttling control valves as
the heating fluid to the feedwater mixing preheater (FWMP) where it additionally heats thewater downstream of the high pressure preheater to the temperature of 355°C at the pressureof 170 kg/cm2. The feedwater preheating at such a temperature is necessary to prevent thelead coolant freezing (Tmelt = 327°С) in the primary circuit in emergency conditions. It is forthe same purpose that cutoff valves (two for each SG) are installed on the pipelines forfeedwater supply to the SGs. They are closed automatically both by the “passive drive” andthe “active drive” in response to respective emergency signals. Feedwater pumps (FWP-2)(two working pumps and one standby pump) supply feedwater from the FWMP to each SG
via control units with the main and starting control valves. The FWP-2 speed is controlled in
the range from 70 to 100% of the nominal value. Each FWP-2 is equipped with arecirculation line. The common header serves as a starting recirculation line to the FWMP
with a capacity of 350 t/h and a control valve for warming up the entire system prior to thewater supply to the SG. The peculiarity of the secondary circuit is that it is not charged withsafety functions for emergency heat removal from the reactor.
5. Safety Features
The principle of natural safety is achieved through the inherent physical and chemical
behaviour and properties of fuel, coolant and other reactor components, negative reactivity
feedbacks, natural dependencies and neutron balance in fast reactors. Accidents are avoided
due to the intrinsic safety features of BREST, including reactivity fuel temperature
coefficient, coolant and core design components, and also coolant pressure and temperature at
the core inlet and outlet. For this reason BREST can be considered an intrinsically safe
reactor. However, the search for the weaknesses in the design of the reactor and its safety296systems in terms of safety continues, and it is not excluded that the list of predominantly
passive engineered features for overcoming accidents caused by previously ignored initial
events will be expanded. An accident with a SG tube rupture is one of the most adverse
events for BREST-OD-300. To reduce the consequences of a potential accident with a SG
tube rupture, a mixed integral/loop configuration of the primary circuit is adopted, with SGs
and MCP installed outside the reactor central vessel. Together with the selected lead
circulation pattern and steam dump from the reactor gas volume to the localization system,
such configuration excludes the ingress of the hazardous steam into the core and reactor
overpressure.

Figure 5: Cross sectional view of the BREST (Reproduced courtesy of NIKIET)
Emergency Core Cooling System
The emergency core cooling system (ECCS) uses field pipes, immersed directly in lead,which may be used to cool down reactor under normal conditions. The system coolant
circulation in emergency heat removal mode is by natural circulation, with the system coolant
under atmospheric pressure. The system consists of 4 loops. The ECCS air circuit inlet air
temperature operates at a minimum and maximum temperature of –55oC and 37oC
respectively. By-pass line incorporated in the coolant loop prevents inadvertent valve
actuation in pumps shut down. The safety analysis has shown that all considered initial events
involving a fast introduction of reactivity up to its full margin (spontaneous movement of all
CPS control rods, steam throw to the core at a SG tube rupture, etc.), interruption of theforced coolant circulation (trip or arrest of all pumps), loss of secondary heat sink or leadsuper cooling at the core inlet and so on, do not lead to accidents with fuel damage andinadmissible radioactive or toxic releases, even in the case of a failure of the reactor active
safety systems.
6. Plant 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 and other structures are mounted on a single monolithic
reinforced concrete foundation plate. In order to reduce seismic inertia forces, the building isdesigned to be symmetrical with the footprint of 65×74 m and separated with intercoupled
vertical bearing diaphragms shown in Figure 2. The main building’s structures have been
designed based on the “Regulations for Aseismic Design of Nuclear Power Plants”. The
297

magnitude of 6 is assumed in the design as the design basis earthquake and the magnitude of
7 on the MSK-64 scale is assumed as the ultimate design-basis earthquake. The buildings are
separated from each other with an aseismic temperature-deformation joint.
Reactor Building
The reactor building and fuel storage area are equipped with a full monitoring system withclosed circuit monitoring system (CCTV).
Balance of Plant:
Turbine Generator Building
The reference concept of the turbine plant has been developed.
Electric Power Systems
These systems include the main generator, main transformer, unit auxiliary transformers,
stand-by auxiliary transformers, diesel generators, and batteries. The electrical systems,
including the class 1E and non-class 1E, are based on a two train approach.
7. Design and Licensing status
The development and construction of the BREST-OD-300 reactor is included in theframework of tasks in:•••

Development Strategy of Nuclear Power in the Russian Federation in the first half ofthe 21st century approved by the Russian Government in 2000;
The Federal Target program “Nuclear Power Technologies of a new generation for theperiod of 2010-2015 and up to the year 2020” approved by the Russian Government
in 2010;The “Proryv” project (2011) that integrates projects on the strategic solution of target
tasks on the creation of natural-safety nuclear power technologies based on fastneutron reactors and a closed nuclear fuel cycle (CNFC).

The BREST-OD-300 power unit is designed as a pilot and demonstration unit intended forstudying the reactor facility operation in different modes and optimizing all processes andsystems that support reactor operation. Furthermore, BREST-OD-300 is considered theprototype of a fleet of medium sized power reactors.
8. Plant Economics
A plant with a BREST-type reactor is expected to be economically competitive owing to thesimpler design of the facility and its safety systems, as well as to efficient utilisation of
nuclear fuel and generated heat. The integral configuration of the circuit components in a
concrete pool predicts a reduced construction costs. On-site fuel cycle arrangement is also
likely to be economically beneficial owing to the shorter out-of-pile cooling andtransportation time, which will eventually lead to a reduction in the recycled fuel quantity –one of the greater contributors to the fuel cycle costs. Comparative economic estimates allow
one to hope to the expenses not higher than298

SVBR-100 (JSC AKME Engineering,Russian Federation)
1. Introduction
The SVBR-100 is a multipurpose small modular fast reactor lead–bismuth (LBE) cooled withan equivalent electric power of 100 MW. In the Russian Federation, lead–bismuth cooled
reactor technology has been used in several nuclear submarines. 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 is based on the experiencegathered in the design and operation of several LBE facilities on nuclear submarines (NSs),
which allows:
•
•
•
•
•
•
•
•

Use of mastered LBE technology;
Use of almost all basic components, units and equipment devices of the reactor
installation, which are verified 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 Po210 radionuclide; andMultiple LBE freezing and unfreezing in the reactor facility.

2. Target Application
The possibility of multi-purpose application of modular nuclear power plants (NPP) ofdifferent capacities (100 – 600 MW) based on SVBR-100 creates the conditions for satisfying
the requirements of consumers in a new sector of regional and small-scale atomic energy
industry: creation of regional NPP and nuclear co-generation plant (NCGP) of low andmedium capacity utilization as part of floating NPPs 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 from the cities;
Refurbish of NPP units whose reactors have expired their lifetime;
Nuclear desalination systems.

3. Development millstones
2010-2016

Research & development and design of the reactor unit and the pilot plant

2016-2019

Pilot plant and supply of equipment

2019

Pilot plant physical and power generation start up

2024-2025

Serial production and supply of packaged equipment

4. General Design Description
The main components and sub-systems of the primary circuit which includes: reactor
monoblock, secondary circuit, passive hear removal system (PHRS) are shown in Figure 1.
All the equipment of the primary circuit are installed inside the vessel of the reactor

299

monoblock (RMB). The removable unit composed of the core, the control rods and ashielding plug is installed in the centre of the RMB and it is surrounded by an in-vessel
radiation shielding (boron carbide) with SG and MCP modules placed inside it as shown in
Figure 2.

PHRS tank

PHRS
pipeline
MFC circuit
separator
MFC circuit
pump

RMB

Gas systemcondenser

Relief tank(bubbler)

Figure 1: The SVBR-100 reactor system configuration(Reproduced courtesy of JSC AKME Engineering)
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 morethan 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 the reactor 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;
Possibility of creation of module based structured NPP with power multiplying byadding the reactors.

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 passes heat toa two-circuit heat-removal system and SG with multiple-circulation secondary coolant
300system. Natural circulation of coolant in the reactor heat removal circuits is sufficient topassively cool down the reactor and prevent hazardous superheating of the core. The coolant
technology system includes mass-exchangers, 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), arewholly realized by components of the in-vessel components, without using pipelines andvalves. Within the main circulation circuit (MCC), the coolant flows according to thefollowing scheme, as shown in Figure 3 (flow diagram). Being heated in the core (crosssection of the core is shown in Figure 4), coolant flows to the inlet of the medium part of theinter-tube chamber of 12 SG modules connected in parallel to each other.

Main circulation
pump (2)Control rod drive
mechanism (50)

Steam generator
modules (12)Reactor monoblock
vesselCore

Guard vessel

Figure 2: Equipment arrangement in the RMB vessel of SVBR-100(Reproduced courtesy of JSC AKME Engineering)
Then coolant is divided into two flows. One flow moves upwards in the inter-tube chamberand enters into the peripheral buffer chamber with a free level of the “cold” coolant. Anotherflow moves downwards and enters into the outlet chamber out of which it goes to thechannels into in-vessel radiation shielding. Coolant flow moves upwards through in-vessel
radiation shielding and cools it, and then it enters into the peripheral buffer chamber as well.
Out of the peripheral buffer chamber, the coolant flow goes over the downcomer circularchannel along the RMB vessel via the inlet chamber to the MCP suction. Out of the MCP thecoolant flows over the two channels installed in the mono-block of the lower zone of invessel radiation shielding into the distributing chamber, from which main part of flow goes tothe reactor inlet chamber, thus closing the MCC circuit. Very small part of flow moves
upwards via the gap near RMB vessel wall, cooling it and goes into the peripheral bufferchamber.

301

Tank of PHSRSG

Dry saturated
steam

Separator
Feed water
Pump

Gas systemcondenser

Protective
deviceReactormonoblock
Reactor vault
BubblerFlooding
Reactor vault

Figure 3: SVBER-100 Flow diagram (Reproduced courtesy of JSC AKME Engineering)
Secondary System
The secondary system includes: SG modules, feedwater and steam pipelines, separators andautonomous cooling condensers. The basic equipment of SVBR-100 is installed in an 11.5 mhigh tight box-containment. In the lower part of each box there is a concrete well to flooded
water in an event of beyond the design accident the failures all four PHRS via SGs. The
reactor monoblock is installed inside concrete well and is fastened on the head ring of roof. In
the upper part of the box there is the reactor equipment, including four steam separators andfour 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 thesecondary circuit in all reactor operation modes. The gas system condensers are installed in
the upper part of the box in the separate concrete compartment. Arrangement RMB in thereactor vault is shown in Figure 5.
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 coreconfiguration allows for a lower power density compared with the nuclear submarines using
LBE reactors. This design has the capability to utilize various fuel cycles. The first stage willbe the typical uranium oxide fuel leading to a core breeding ratio (CBR) of 0.84; MOX fuelcan also be used, leading to a CBR just about 1. Using UO2 as the starting fuel, the closed
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 in the initial core. The reactor is designed to operate foreight years without core refuelling.

302

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer
Country of origin

JSC Institute for Physics and Power Engineering and JSC EDB
id
RussianFederation

Reactor type

Liquid metal cooled fast reactor

Electrical capacity (MW(e))

100Thermal capacity (MW(th))

280

Expected capacity factor (%)< 90

Design life (years)

60

Coolant/moderator

Lead-Bismuth eutectic alloy

Primary circulation

Forced circulation

System pressure (MPa)Low pressureCore inlet/exit temperatures (oC)340 / 485

Main reactivity control mechanism

Control rod drive mechanism

RPV height (m)7.55

RPV diameter (m)4.53

RPV or module weight (metric ton)

280

Configuration of reactor coolant system

Integral type

Power conversion process

Indirect Rankine cycle

Fuel type/assembly array

UO2

Fuel assembly active length (m)2.1

Number of fuel assemblies

61

Fuel enrichment (%)< 19.3

Fuel burnup (GWd/ton)

60 (average)

Fuel cycle (years)

7–8

Cogeneration capability

Yes

Approach to engineered safety systems

Passive

Number of safety trains

4

Refuelling outage (days)

60 days in 7-8 years

Distinguishing features

Modules per plantIntegral (monoblock) layout of the primary circuit where allcomponents (reactor, steam generation modules, main circulating
pumps) are installed arranged in one vessel
1–6

Target construction duration (months)

36

Seismic design (g)0.5Core damage frequency (per reactor-year)

1E–8

Design Status

Detailed design for potential construction in early 2020s

303

Emergency protection rods (6)Additional emergency
protection rods (12)

FSA without rods (7)Automatic control rods (2)Controlchannels

Supportingtruss

Compensating rods (22)Control and compensatingrods (12)Reactormonoblock(RMB)

Equipment ofreactor cavity

Figure 4: Cross section of the SVBR-100 core
(Reproduced courtesy of JSC AKME Engineering)

Figure 5: Reactor monoblock in thereactor vault(Reproduced courtesy of JSC AKMEEngineering)

5. Safety Features
Physical basement for high level of inherent self-protection and passive safety:
•

•
•
•

First, this is potential energy contained in coolant. LBE does not contain potential
energy, which in an event of accident occurrence can cause destruction of defense
barriers, core damage and disastrous exhaust of radioactivity. This is potential energy of
coolant compression and potential chemical energy of coolant’s interaction with sodium
coolant structural materials (zirconium), water and air;Potential energy is a natural property of coolant and cannot be changed by any technicalsolutions;
Further, it is an integral structure of the reactor facility that completely eliminates
pipelines and valves with radioactive coolant and eliminates an opportunity 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, negative void
reactivity effect. Efficiency of the strongest absorbing rod does not exceed 1 $, that been
coupled with technical performance of the control and protection system exclude an
opportunity of prompt neutrons runaway in the reactor.

SVBR-100 safety does not depend on the state of the systems and equipment of the turbinegenerator installation. Distinguishing features of the SVBR-100 – Safety concept is shown in
Figure 6.

304

Chemically inert
lead-bismuth coolant

Inherent (by-nature)safety – for free
Stability & simplicity
under normal operation+
Fast reactor

First circuit
low pressure

Integral design of the reactor
components:
core, pumps, SG, etc.

By-design
safetyTolerance to the design
andbeyond-design basis
accidents

Figure 6: Distinguishing Features of the SVBR-100 – Safety Concept(Reproduced courtesy of JSC AKME Engineering)
6. Instrumentation and Control systems
The main principles of the design include:
•
•
•
•
•
•

Distributed I&C system with several level of hierarchy and defense in depth;
Soft control of NPP technological systems;
Availability of the large screen and reserve zone at main control room (MCR);
Principles of diversity, reliability, physical separation and others, providing high levelof functional reliability, including protection against common cause failures;
Well-developed diagnostic functions;Self-diagnostic of I&C programmable devices.

Design specificities are:
•
•
•
•
•
•

Regulation of coolant flow rate by changing rotation speed of MCP depending onreactor power for maintaining constant coolant heat up;
Full scope diagnostic system of NPP;
Providing I&C operability during 7-8 years of continuous NPP operation;
New tasks of neutron flux monitoring while core refuelling;Highly reliable reactor control at start up and operation;
Load-follow operation in the deep range (100–50–100%).

7. Plant Arrangement
With the purpose of using SVBR-100 modules in large-scale NPPs, a Russian consortium of
IPPE, Gidropress, FSUE Atomenergoproekt and others have developed a conceptual designof a NPP which consists of two power units of 1600 MW(e) each. A NSSS consisting of 16
SVBR-100 reactor modules and a turbine of 1600 MW(e) in each power unit is arranged in
the main building. A modular design of the NSSS power unit allows have load factor (LF) notless than 90% under long reactor operation. When a reactor module is shut down forrefuelling, power of the unit reduces slightly. Modular NPPs also allow fabrication of thesingle modules in a factory, easy water or ground transportation, include rail way, of themodules to the NPP site and reduced construction times. General view of pilot plant is shownin Figure 7.

305

Figure 7: Pilot plant of the NPP with the SVBR-100 reactor module(Reproduced courtesy of JSC AKME Engineering)Control building
Control building will be developed later.
Balance of plant:
Turbine Generator building
There are individual turbines for each of the reactor modules that are skid mounted andstandard models currently available or several reactor modules can provide of steam
one turbine-generator.
Electric power systems
Electric power system for pilot plant of 100 MW capacity is underway.
8. Design and Licensing Status
The Rosatom Scientific and Technical Council convened on 15 June 2006 approved thedevelopment of the technical design of experimental industrial power unit based on the
SVBR-100. Siting licence is received to current time at Dimitrovgrad, in the region of
Ulyanovsk. Key reactor and reactor core research and development works have begun. A
complete reactor and power plant design is expected to be completed by 2020, along with apreliminary safety report. A construction licence is also expected to be obtained. The trial unit
is expected to be commissioned in early 2020s.
9. Plant Economics
Russian Federation built seven Alfa-class submarines, each powered by a compact 155
MW(th) LBE cooled reactor, and 80 reactor-years operational experience was acquired withthese. Overnight capital cost for pilot plant is previously estimated as $4000-4500/kW andgenerating costs 4-5 c/kWh on 90% load factor. Technical and economical assessments haveshown that a number of SVBR-100 modules can replace the 2nd, 3rd, and 4th unit of the
Novovoronezh NPP (NVNPP) at half the capital cost of replacing the same power capacity.
The SVBR-100 structural design includes the requirements of versatility in terms of the fuelused to enable fast transition to MOX and later to nitride fuel as well, to serve the basis forfuel self-sufficiency in the closed nuclear fuel cycle.

306

G4M (Gen4 Energy Inc., United States ofAmerica)
1. Introduction
The G4M power module is a low capacity, portable and self-contained reactor designed todeliver a 70 MW(th) power with lead–bismuth eutectic as the primary coolant. The G4M
does not need a human operator to generate a capacity of 25 MW(e). It has a fuel cycle of 10
years without on-site refuelling. The G4M conceptual design is being developed by the Gen4
Module (G4M) Inc. formerly Hyperion Power Generation Inc. formed in 2007. The design
aims for achieving the operations of the standalone small power module simple enough that
only a limited number of personnel are required to keep it operating safely and reliably. The
reactor is intended to be sealed at factory, sited underground and eventually returned to thefactory for fuel recycling and refuelling after a useful life of 10 to 15 years. The G4M has thefollowing attributes:• Single-unit, sealed construction and underground siting;• Inherent simplicity and compactness of the design enable mass production and ease ofshipping in off-the-shelf containers;• Modest size and design simplicity greatly reduce the financial investment and enhancesystem reliability;• Low power output and passive backup coolant systems combine to provide a highdegree of safety.
Figure 1 shows the G4M reactor module based on 25 MW(e) power plant.

Figure 1: The G4M reactor module based on 25 MW(e) power plant(Reproduced courtesy of Gen4 Energy)
2. Target Application
The flexibility of G4M module to produce steam/electricity lends itself to applications
ranging from supplying process heat for heavy industry to electricity for consumers electrical
power applications. The design organization identified three main markets for which its
307technology has a potential to be deployed: mining/oil and gas production, isolated and island
communities and government facilities. Although G4M design can be located in any
environment, the design is strategically best suited for remote locations or harsh climate
environments.
3. Development Milestones
2007

Reactor design started

2008–2009Engineering specifications

2010

Plans to build a G4M at Savannah river site in partnership with Hyperion Power
Generation (HPG)
Announced that they would not be pursuing the US Department of Energy’s SMR
licensing support programmeConceptual design completed

2012
2015

4. General Design Description
Design Philosophy
The G4M is designed to avoid the high construction costs and long construction times of the
traditional large light water reactors and to provide power to remote areas. Another design
objective of the G4M reactor design strategy is transportability and safety during that
operation. The reactor concept is designed to eliminate or reduce the likelihood of accidents
during transportation. This includes:•••

Designed to eliminate the potential of criticality accidents;
Elimination of any interaction of the reactor coolant with water (by choice of leadbismuth eutectic as a reactor coolant);
Protection from fire and a highly reliable means of cooling on transport back to thefactory.

The G4M power module cycle is shown in Figure 2.

Figure 2: The G4M power module cycle (Reproduced courtesy of Gen4 Energy)

308

Reactor Core
The reactor is approximately 1.5 m in diameter and 2.5 m in height, in which there are 24
subassemblies containing the fuel pins. The adopted fuel is uranium nitride (UN) pellets with19.75% enrichment is contained in clad tubes made of HT-9. Quartz is used as a radialreflector. These high-temperature ceramic material pellets deter the ability to separate
plutonium from spent fuel. The pin assembly is filled with liquid LBE to provide a highconductivity thermal bond between the fuel and cladding. The gap in the fuel pins is sized topreclude fuel clad mechanical interference throughout the core’s lifetime. A plenum is locatedat one end, which serves both as a fission gas plenum and a repository for the LBE inside thepin as the fuel swells with burn-up. The fuel pellet stack is between two and three meters
high, which includes neutron reflectors and the gas plenum. There are four types of subassembly configurations shown in Figure 3. The geometries of these sub-assemblies are
dictated by the locations and sizes of the reserve shutdown cavity and control rods in the corearray.
Reactor vessel
Sub-assembly (24)
Shutdown Control Rod
Operational
Control Rod
Cavity for Reserve
Shutdown Balls

Figure 3: G4M core plan view (Reproduced courtesy of Gen4 Energy)
Reactivity Control
There are three (3) independent reactivity shut-down systems in the core: (1) a shutdown rod
system composed of six (6) boron carbide (B4C) rods; (2) a control rod system comprising 12boron carbide (B4C) rods; and (3) a reserve shutdown system consisting of a central cavity
into which B4C balls may be inserted. Each of the three systems can independently take thecore to long-term cold shutdown. The rod shutdown and the ball shutdown systems perform
this safety function automatically and instantaneously when triggered. Both the control rods
and the spheres are inserted into dry wells in the core, which are hexagonally shaped
thimbles. These thimbles penetrate the reactor vessel and are sealed from the primary coolant.
Both systems can independently take the core to long term cold shutdown.
Reactor Pressure Vessel and Internals
The outer diameter of the reactor system, including the outer reflector and coolant downcomer, is limited to 1.6 m to be able to seal the reactor vessel system at the fabrication facility
and transport it to the site in a conventional shipping cask. The total mass of the reactor
vessel with fuel and coolant is 18 metric tonnes.

309

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Gen4 Energy Inc.

Country of origin

United States of America

Reactor type

Liquid metal cooled reactor

Electrical capacity (MW(e))

25Thermal capacity (MW(th))

70

Expected capacity factor (%)> 90

Design life (years)

5–15 (nominal 10) years

2

Plant footprint (m )–Coolant/moderator

Lead-bismuth

Primary circulation

Forced circulation

System pressure (MPa)Low pressure operationoCore inlet/exit temperatures ( C)– / 500

Main reactivity control mechanism

Rod insertion and B4C ball insertion

Power module height (m)2.5Power module wide (m)1.5RPV or module weight (metric ton)

18

Configuration of reactor coolant system

Integral pool type

Power conversion process

Indirect Rankine cycle

Fuel type/assembly array

Uranium nitride

Fuel assembly active length (m)-Number of fuel assemblies

24

Fuel enrichment (%)

19.75

Fuel burnup (GWd/ton)–

Fuel cycle (years)

10

Cogeneration capability

Yes

Approach to engineered safety systems

Hybrid (Active + Passive)

Number of safety trains

2

Refuelling outage (days)–

Distinguishing features
Modules per plant

Transportable factory fuelled design, a fuel cycle of 10 years,
under grade containment system.
1

Target construction duration (months)–

Seismic design–Core damage frequency (per reactor-year)–

Design Status

Conceptual design

Reactor Coolant System
The core coolant is LBE, with a mixed mean exit temperature of 500°C. This temperature
limits the cladding temperature, so that maximum cladding creep over the 10 year lifetime of
the reactor is less than 1%. A solid phase oxygen control system used to control the oxygen
level in the coolant to maintain a protective coating on the clad surfaces, limiting corrosion in
the clad to acceptable levels. The pitch-to-diameter ratio of the fuel pins in the core was set tolimit the coolant velocity to 1.9 m/s, to prevent erosion of the protective coating on the fuel310

pin clad surfaces. The use of LBE liquid metal coolant enhances the safety and radicallyreduces the potential energy available for dispersion of radioactivity in a worst-case accident.
•••

Potential radioactivity releases from vaporization of the coolant are eliminatedbecause of the high boiling temperature of LBE;
Potential radioactivity releases due to pressurized release of the coolant are eliminatedas LBE is at near-atmospheric pressure;
LBE is non-reactive with air, water, fuel and metal, thus rapid exothermic reactions
with associated radioactivity releases are eliminated. Additionally, reactions that
produce flammable hydrogen are eliminated.

The heat exchanger placed on the top of the sealed vessel transfers the thermal power
generated in the reactor core to the secondary heat pipes that are larger and longer so they canreach grade level and subsequently transfer the thermal power to a working fluid used as ahear supply or used to drive rotating machinery to produce electricity.
5. Safety Features
The several safety features of the G4M module result in a safe, reliable and proliferationresistant method of generating power, that is analysed thoroughly for all radioactive related
releases, natural threats such as seismic, tornado/missile, etc. Among the safety principles thatG4M power module has adhered to include:•••

Natural circulation of the primary coolant to remove the heat from the core in theevent of a loss-of-flow incident;
Using highly reliable passive safety features to decrease the probability of failure;
Using the principle of defense-in-depth to further decrease the frequency of accidents
that could impact public safety, including using multiple containment to prevent any
release of radioactive materials. This principle also applies to physical security as an
objective to prevent malevolent intrusion from any outside threat.

Decay heat removal system
During operational shutdown, decay heat is removed from the G4M by two methods. During
operational shutdowns, decay heat is removed from the core through the normal coolant
pathway by dumping steam to the condenser. The second backup decay heat removal systemprovides natural circulation of LBE through a fixed bypass path in the core. Water from anemergency cooling tank is gravity-sprayed onto the exterior surface of the G4M module, andheat is removed by passive vaporization of water. The backup decay heat removal system canprovide adequate cooling for up to 14 days without any power or operator action.
Containment system
The G4M module is sited in an underground containment vault to provide isolation from theenvironment, prevent intrusion or tampering and avoid accidents from extreme external
events. The inner vessel confines the hydrogen gas and provides containment for all
radioactive materials. The containment vessel is designed for double confinement. The
hardened containment vault is designed to protect against extreme external events and manmade events. Any radioactivity release from the G4M module, however unlikely, is containedin the vault. The underground vault eliminates or minimizes the potential impact of natural
disasters such as floods, hurricanes and tornados by preventing any contact of wind, water or
debris with the reactor. The vault prevents the possibility of intrusion or tampering, as well as
minimizing the impact of security threat scenarios such as an airplane impact or an explosive
device. Because the G4M module is factory sealed, never opened on site, and located in an311

underground vault, the potential for any radioactive contamination on the site is virtuallyeliminated.
6. Plant safety and Operational Performances
The reactor operates at an optimum temperature of 550°C, picked as the goal for the
Generation IV reactors. At 550°C, the dissociation pressure for the hydrogen above thehydride is approximately eight atmospheres, which permits easy transportation of the gas
without presenting significant high-pressure risk. The temperature-driven mobility of thehydrogen contained in the hydride can change the moderation, and therefore the reactor
criticality, making the reactor self-regulating. The hydrogen forced out of the core during any
over-temperature excursion reduces the neutron energy moderation necessary for nuclear
criticality. The G4M power module is inherently fail-safe, since any temperature increase
from excess activity immediately reduces the criticality parameters and thus the power
production. The consequent power reduction causes the temperature to decrease and thattemperature decrease eventually reverses the process, resulting in relaxation oscillations thatquickly damp out to steady-state operation.
7. Instrumentation and Control systems
The complex arrays of detectors, analysers and control systems responsible for the safety andstability of conventional nuclear power reactors have been superseded by the fundamental
science and properties of the active materials.
8. Plant Arrangement
The G4M plant layout diagram is shown in Figure 4, below.

Figure 4: The G4M plant layout diagram (Reproduced courtesy of Gen4 Energy)312

9. Design and Licensing Status
Gen4 Energy presented its conceptual design to the U.S. NRC, and according to the company,
to appropriate licensing authorities in other countries. The company believes that its initialclients will be best served by licensing the design either in the U.S., Canada or the U.K.
Although the first unit is likely to be deployed at the Savana River Site, Gen4 Energy is also
in talk for potential deployment of the G4M in Canada. Gen4 Energy announced in April
2012 that they would not be pursuing the US Department of Energy’s small modular reactor
licensing support programme because they concluded that, use of well-known light water
reactor technology of 45 to 300 MW(e) intended for deployment in the United States had amuch higher probability of success given the (funding opportunity announcement’s) statedmaximum of two awards.
10. Plant Economics
Power generation is expected to be economically better than the large-scale nuclear power
plants and cheaper, from a broad point of view, than coal, gas, fuel oil, etc.

313

EM2 (General Atomics, United States ofAmerica)
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). An elevation view of an EM2 modular building element is shown in
Figure 1. The reactor employs a “convert and burn” core design which 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 an R-245fa Rankine bottoming
cycle for 53% net power conversion efficiency assuming evaporative cooling. EM2 is multifuel capable, but the presented case assumes low-enriched uranium (LEU) with depleted
uranium (DU) fertile fuel.

Figure 1: Elevation view of EM2 modular building element employs two modules on a single
seismically isolated platform (Reproduced courtesy of General Atomics)
2. Target Applications
The EM2 is being developed for the electricity generation and high temperature use.
3. Development Milestones
Phase I

High risk R&D
Conceptual design of reactor core and power plant
Fuel design and demonstration
High temperature material development

Phase II

Demonstration/prototype plant
Qualification of fuelQualification of plant operation

Phase IIIFirst-of-a-kind plant
Demonstration of commercial operation

314

4. General Design Description
Design Philosophy
EM2 design initiative is to develop a new nuclear plant to address the following goals forenhancing the likelihood of commercial success:••••Economic parity with fossil fuel generation in the U.S.
Improved siting flexibility via dry cooling and site accessibility
Passive safety for sustained protection during long-term station blackout and othersevere accidents
Improved fuel resource utilization, reduced nuclear waste and high proliferationresistance

Reactor Core
The core is supported by the support floor throughthe core barrel, which attaches to the vessel below
the cross-duct. The upper carbon-composite (C-C)
heat shield protects the top head elements fromthe hot helium. The vessel is internally insulatedwith silica/alumina fibrous insulation retainedwith a C-C cover plate. In order to achieve highfuel utilization, the EM2 core utilizes the “convert
and burn” concept, in which the core is divided
into two core sections: fissile and fertile. The
fissile section is the “critical” section at beginningof life (BOL). It contains ~14.5% LEU to sustain
the chain reaction and provide excess neutrons toconvert DU fertile to fissile material. The averageenrichment of the total active core is 7.7%. The
reflector consists of an inner section of zirconiumalloy blocks and an outer section of graphite
blocks.

Figure 2: Reactor core arrangement(Reproduced courtesy of General
Atomics)

The basic building block of the advanced EM2 fuel system is the hex-assembly, of which
there are 85 in the core. Eighty-one hex-assemblies are joined into 27 tri-bundles and 4
remain as individual hex assemblies. The tri-bundle is located between separate upper andlower reflector assemblies. It has a bottom alignment grid, an upper manifold assembly, andone intermediate spacer grids. The fuel for EM2 is contained in cylindrical fuel rods arranged
in a triangular pitch. Due to the high operating temperatures and long fuel cycle, all tri-bundle
structural components and cladding are made of SiC-SiC.
Fuel Characteristics
Uranium carbide (UC) is used to meet the high uranium loading requirement; has a very highthermal conductivity; is compatible with the SiC-SiC cladding; and has a suitably highmelting point. Each annular fuel pellet is a sintered “sphere-pac” with a specified interstitialand internal distributed porosity to allow for faster migration of volatile fission products. SiCSiC 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.

315

MAJOR TECHNICAL PARAMETERS:Parameter

Value

Technology Developer

General Atomics

Country of Origin

USA

Reactor Type

Modular High Temperature Gas-cooled Fast Reactor

Electrical Capacity (MW(e))

265

Thermal Capacity (MW(th))

500

Expected Capacity Factor (%)

90

Design Life (years)

60

2

1

Plant Footprint (m )~90000 (4 modules)

Coolant/Moderator

Helium / None

Primary Circulation

Forced circulation

System Pressure (MPa)13.3 (peak)

Main Reactivity Control Mechanism

Control rod drive mechanism

RPV Height (m)12.5

RPV Diameter (m)4.6

Coolant Temperature, Core Outlet (°C)

850

Coolant Temperature, Core Inlet (°C)

550

Integral Design

No

Power Conversion Process

Direct Brayton top cycle and Rankine bottom cycle

High-Temp Process HeatNo

Low-Temp Process HeatNo

Cogeneration Capability

Yes

Design Configured for Process Heat Applications

No

Passive Safety Features

Direct reactor auxiliary cooling system (DRACS)

Active Safety Features

Control rods and shutdown rods

Fuel Type/Assembly Array

UC pellet/ hexagon

Fuel Enrichment (%)~14.5

Fuel Burnup (GWd/ton)~130

Fuel Cycle (months)~360

Number of Safety Trains

N/A

Emergency Safety Systems

Control rods and shutdown rods

Residual Heat Removal Systems

Passive

Refuelling Outage (days)

14

Distinguishing Features

Silicon carbide composite cladding and fission gas vent system

Modules per Plant

4

Estimated Construction Schedule (months)

42

Seismic design (g)N/APredicted large release frequency

N/A

Design Status

Conceptual design316Fuel Handling
The core is accessed by a refuelling machine from the maintenance hall floor. An articulatedarm 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 thesealed containers by passive natural convection of air. No water or active cooling is required.
Reactivity Control
Reactivity control is provided by the 18 control rods and 12 shutdown rods. Both control rod
system and shutdown rod system each have sufficient negative reactivity to render the core
cold subcritical. The control and shutdown drives are located at the top of the reactor vessel.
The control rod drives utilize a ball-screw drive while the shutdown rods use linear motors.
Reactor Pressure Vessel and Internals
The reactor pressure vessel (RPV) is constructed fromwelded ring-forgings or rolled plate. The bottom headcontains three large penetrations for the two cross-vessels
and a flanged hatch near the bottom. The top head has thepenetration for refuelling access 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, whichallows the use of SA-533 grade B material.
Power Conversion Unit
The power conversion is based on a combined cycle witha direct helium Brayton cycle and a Rankine bottoming
cycle. The helium Brayton cycle is located in the PCU
vessel, while Rankine cycle is located outside the reactor
building. The Brayton cycle incorporates the turbocompressor (T/C) and generator which are mounted on an
in-line vertical shaft suspended by active magnetic
bearings as shown in Figure 3. The cycle also incorporates
two heat exchangers (HX), a recuperator and a precooler.
The generator uses a permanent magnet (PM) rotor toeliminate I2R losses associated with a wound rotor andexciter.

Figure 3: EM 2 PCU enclosed
Brayton top cycle(Reproduced courtesy ofGeneral Atomics)

5. Plant Safety Features
Description of Safety Concept
The EM2 safety design uses a defence-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 firstbarrier is the SiC-SiC fuel cladding. The second barrier is the primary vessel system, which
encompasses the reactor, PCU, and DRACS. The third barrier is the free-standing, belowgrade containment. Because the fuel is vented, the fission product vent system (FPVS) is anextension of the first barrier.

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Fuel Cladding
The SiC-based composite cladding is a superior material being developed for accident
tolerant fuel (ATF). SiC-SiC maintains its strength up to 1700ºC and retains greater than
75% of its strength at 2000 ºC.
Fission Product Vent System
The FPVS protects fuel cladding by venting gaseous fission products to high temperatureadsorber in the containment. The venting effectively reduces the volatile fission productinventory in the fuel rods and reduces the source term for accidents.
Direct Reactor Auxiliary Cooling System
The DARCS safely removes core afterheat during normal shutdown and accident conditions
when the PCU is not available. It provides controlled core heat removal during an anticipatedtransient without scram (ATWS) and active core heat removal during special maintenanceconditions characterized by need for low temperatures and/or low helium pressures.
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 containmentis hermetically sealed with an inert (argon) atmosphere at ~20 psig. The peak pressure rating
is 90 psig. The design leakage rate is less than 0.2% per day.
Protection from External EventsProtection from external events is provided by locating the reactor containments and thespent fuel storage facility below grade as shown in Figure 4. The roof of the maintenance hall
above the reactors and spent fuel storage is an aircraft crash shield per NRC regulations. The
containment and reactor auxiliary building are mounted on a common seismic isolation
platform similar to that used for large building in seismic areas.

Figure 4: Vertical cross-section of EM 2 plant showing below-grade containment andcrash shield. (Reproduced courtesy of General Atomics)

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6. Plant 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 loaddemand. In the automatic load following mode, the generator speed is set by modulating thegenerator speed. A field-oriented control algorithm in the frequency converter controls thegenerator torque that decreases or increases the generator speed, which in turn determinedhelium 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 constanttemperature with load so that rapid load following should be possible.
7. Instrumentation and Control Systems
EM2 deploys advanced sensors for high operating temperature condition such as solid-stateand silicon carbide (SiC) neutron flux monitors. The plant control including startup, operationand shutdown is conducted through integrated control system actions, which regulate reactor
power and turbo-machinery 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 system(RCS). The process component control system includes a non-synchronous, variable speed
generator and a frequency inverter.
8. Plant Arrangement
The baseline EM2 plant is composed of four 265 MW(e) modules for a combine net power of
1,060 MW(e) to a 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 themodules can be built sequentially and operated independently. Figure 5 shows the plantlayout, which 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.

Figure 5: One gigawatt EM 2 plant layout on 9 hectares of land (Reproduced courtesy ofGeneral Atomics)

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Reactor Building
The reactor building as shown in Figure 1 is divided into two sets of two module separated
by the electrical distribution building and access entry. Two reactor modules with individual
containment assemblies are mounted on a seismic isolation platform. The reactor auxiliarybuilding is also mounted on the platform. The containment structure is suspended from anapproximate mid-plane support frame that also supports the primary system. Access to thereactor, PCU and DRACS units is from the maintenance floor at grade level.
Balance of Plant
The primary system is enclosed within a sealed, containment as shown in Figure 6 with two
chambers connected by a duct. The primary system includes the reactor and PCU whose
respective vessels are connected by a concentric cross-duct. The reactor vessel is also
connected to two DRACS systems (only one shown in Figure 6). Natural circulation paths
are provided by the vertical concentric cross-ducts to helium-to-water heat exchangers. The
maintenance circulators, which are normally valve-off, are used for low pressure maintenanceconditions.

Figure 6: EM2 primary system enclosed in two-chamber sealed containment (Reproduced
courtesy of General Atomics)
9. Design and Licensing Status
The EM2 prototype plant will be licensed using the two-step 10 CFR Part 50 and follow-on
commercial plants will be licensed under the one-step 10 CFR Part 52 process.
10. Plant Economics
The target equilibrium plant overnight capital unit cost of a four module installation is about$4,300/kWe (2014$) plus and additional first core cost of $240/kWe. The equilibrium
levelized cost of electricity is about $67/MWh assuming a weighted average cost (WAC) of
capital of 8.4%.
For further information: http://www.ga.com

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MOLTEN SALTSMALL MODULAR REACTORS322

Integral Molten Salt Reactor (Terrestrial
Energy, Canada)
1. Introduction
The Integral Molten Salt Reactor – 400 Megawatts-thermal (IMSR400) is a small modular
molten salt fuelled reactor [1]. It features a completely sealed reactor vessel with integrated
pumps, heat exchangers and shutdown rods. This vessel, called the IMSR core-unit, isreplaced completely as a single unit 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.
Primary Pumps

Primary Heat Exchangers

Graphite Moderator

IMSR Core-Unit
Guard Vessel
Silo

Figure 1: Core-unit and guard vessel in one of the two reactor silos(Reproduced courtesy of Terrestrial Energy)
For ultimate safety, 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. This is achieved through a combination ofdesign features:• The inert, stable properties of the salt;• An inherently stable nuclear core;• Fully passive backup core and containment cooling systems;• An integral reactor architecture.
2. Target Application
The IMSR plant is designed to accommodate various load users, from baseload to loadfollowing. Featuring a simple modular and replaceable core-unit, very high reliability ratings
are targeted. The IMSR has been specifically designed for factory fabrication. Nuclear
components are small and road-transportable. The IMSR core-unit is designed for a shortservice lifetime, which allows dedicated factory lines to produce the units semiautomatically, similar to aircraft jet engine production lines for example.
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3. Development Milestones
2015

Concept design completed

2016

Start of pre-licencing vendor design review (by Canadian nuclear regulator)

2016

Start of basic engineering phase

Early 2020’s

Secure necessary licenses

Early 2020’s

Plan construction of a first full-scale NPP in Canada

4. General Design Description
Design Philosophy
The design is based heavily on the intensive molten salt reactor project of Oak Ridge
National Laboratory (ORNL) in the 1950’s - 1970’s. In this project, extensive R&D was done
and molten salt reactor materials and equipment were developed. This included such items as
suitable graphite to act as neutron moderator, drain tanks for the salts, hot cells for housing
the equipment, pumps and heat exchangers. The ORNL program culminated in theconstruction and successful operation of the Molten Salt Reactor Experiment (MSRE). The
MSRE employed a molten fluoride fuel salt, which was also the primary coolant. This fuelcoolant mixture was circulated between a critical graphite moderated core and external heatexchangers. The IMSR is essentially a re-configured, scaled-up version of the MSRE. This
minimizes required research and development (R&D). The IMSR has also been influenced bya more recent ORNL design, the small modular advanced high temperature reactor(SmAHTR) which was to use solid fuel with fluoride salt coolant, but had an “integral”
architecture where all primary pumps, heat exchangers and control rods are integrated inside
the sealed reactor vessel.
Power Conversion System

Figure 2: IMSR400 main heat transport path for power generation(Reproduced courtesy of Terrestrial Energy)

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The IMSR uses molten fluoride salt, a highly stable, inert liquid with robust coolant
properties and high intrinsic radionuclide retention properties, for its primary fuel salt. A
secondary, coolant salt loop, also using a fluoride salt (but without fuel), transfers heat awayfrom the primary heat exchangers integrated inside the core-unit. As shown in the Figure 2,
the coolant salt loop, in turn, transfers its heat load to a solar salt loop, which is pumped outof the nuclear island to a separate building where it either heats steam generators thatgenerate superheated steam for power generation or is used for process heat applications.
Reactor and Core-unit
The IMSR is an integral nuclear reactor design. This means that the reactor core, pumps, heatexchangers and control rods are all mounted inside a single vessel, which is referred to as the
IMSR core-unit. This core-unit is manufactured in a controlled factory environment and thenbrought to the reactor power plant site where, following final assembly, it is lowered into asurrounding guard (containment) vessel which itself sits in a below grade reactor silo. There,
the core-unit is connected to secondary piping, which contains a non-radioactive coolant salt.
For optimal safety and containment, this coolant salt is used as an inert, low pressureintermediate loop between the critical reactor circuit and the steam power generation circuit.
The steam circuit powers a conventional, off-the-shelf industrial steam turbine for power
generation and/or industrial steam production, depending on the required application.
Alternately, some or all of the hot molten salt may be sent directly to a process heatapplication.
Fuel Characteristics
Being a liquid fuel reactor, there are no fuel elements in the reactor core. The fuel may
consist of low enriched uranium fluoride, plutonium fluoride, thorium fluoride, or any
mixture of these. The first of a kind IMSR400 however, will utilize a once-through, lowenriched uranium fuel cycle as this is the simplest option. This fuel salt is diluted with
coolant salt, consisting of fluorides such as sodium fluoride, beryllium fluoride and/or lithium
fluoride. Together this mixture constitutes both fuel and primary coolant. The fuel-coolant
mixture is pumped between a critical, graphite moderated (thermal spectrum) core, and thenthrough the integral heat exchangers to transfer its heat to the secondary, coolant salt loop.
The latter loop consists of bare diluent salts (without fuel salt added), that in turn transfers itsheat to another intermediate salt loop. This second intermediate salt loop or “third loop”
improves safety by adding another barrier between the radioactive primary inventory and thesteam turbine. This third loop uses a nitrate salt, for its lower melting point (to avoid steam
generator freezing) and compatibility with steam (in the event of a leak). Finally, the nitrate
salt heated steam generator produces steam for process heat or to power a steam turbinegenerator set.
Fuel Handling and Core-unit Replacement Approach
The new or “fresh” fuel is separately brought to the power plant site as a solid, where it ismelted and added to the IMSR core-unit. This allows the IMSR to operate with online
fuelling. Additionally, and unlike solid-fuel reactors, there is no need to remove a proportionof old fuel during makeup fuelling. All of the fuel stays inside the closed IMSR core-unit
during the entire power operations period of the core-unit. The volume of small amounts ofadditional “makeup” fuel salt is simply accommodated in the upper gas plenum.
Unlike other power reactor systems, the IMSR core-unit needs never be opened at the power
plant site, either during start-up fuelling or during refuelling. After ~7 years of power
operation, the core-unit is shut down and after a cool-down period, the used fuel charge ispumped out to robust holding tanks located inside the containment.

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MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Terrestrial EnergyCountry of origin

CanadaReactor type

Molten Salt Reactor

Electrical capacity (MW(e))

185 – 192

Thermal capacity (MW(th))

400

Expected capacity factor (%)

90

Design life (years)

60

Coolant/moderator

Fluoride fuel salt and graphite moderator

Primary circulation

Forced circulation: 5400 kg/s

System pressure (MPa(a))< 0.4 (hydrostatic)oCore inlet temperatures ( C)625 – 660

Core exit temperatures (oC)

670 – 700oMean temperature rise across core ( C)75

Main reactivity control mechanism

Short term: Negative temperature coefficient; inherent
Long term: Online liquid fuel additions

RPV inner diameter (m)3.5

RPV height (m)7.0

RPV mass (t)170 (Transport weight)

Power conversion process

Supercritical steam turbine; two intermediate salt loops.

Cogeneration / Process Heat Capabilities

District heating, hydrogenproduction, liquid fuelproduction, Industrial co-gen, ammonia production,mineral resource extraction, petrochemical refining

Passive Safety Features:

negative temperature coefficients, large heat capacity, heatremoval by natural means; flow driven control rods

Moderator type

Graphite

Fuel enrichment (%)

Depends on type of cycle, first of a kind to be low enricheduranium

Fuel cycle (months)

84 months before core-unit replacement

Residual Heat Removal System

based on heat large capacity of salt and graphite, heat loss
through the core-unit; guard vessel and surrounding air

Refuelling outage (days)

Not applicable; second unit can be operated

Distinguishing featurescompletely sealed reactor vessel with integrated pumps, heatexchangers and shutdown rods; core-unit is replacedcompletely as a single unit at the end of its useful service

Modules per plant

Dependent on market and application

Target construction duration (months)

Factory manufactured units will reduce on-site construction

Seismic design~0.3g (to be determined in detailed design phase)Core damage frequency (per reactor-year)

Not applicable

Design Status

Conceptual design compete – Basic engineering in progress

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The spent, empty core-unit is then allowed to cool down for several more years, at which
point its radioactivity has diminished sufficiently to permit its safe removal from its
containment silo. The IMSR400 is a single unit plant, however two core-unit silos are used,
to allow switch-loading of the silos: this allows a long period of cool-down for the spent silo,
while another is connected by switching over the secondary, coolant salt lines to the new unit.
After this long cool-down period, the spent, empty core-unit is then moved into a long-term
storage silo where it can stay for an extended period. Beyond this period the core-unit can be
shipped to a central facility, to be recycled or prepared for geological sequestration. Similarly
the separately stored spent fuel charge can also be shipped to a central facility to recover thefuel or for conversion to a form for sequestration. This sealed, integral reactor architecture
and once-through, “burner” fuel cycle mode allows the actual power plant site to always
operate in a clean, simple environment without risk of facility contamination.
The IMSR has intrinsic, design, and procedural features that resist proliferation of fissile
materials for weapons production
Cooling System
The unique cooling system is based on heat capacity and heat loss – which are immutable.
Heat capacity is due to the thermal mass of the fuel salt, vessel metal, and graphite. Heat loss
occurs as the reactor vessel is not insulated. Short term cooling is provided by the low power
density core-unit, and the internal natural circulation capability of the fluoride fuel salt,
resulting in a large capacity to absorb transient and decay heat generation.

Figure 3: The heat loss approach for the IMSR core-unit(Reproduced courtesy of Terrestrial Energy)
Longer term cooling is provided by heat loss from the uninsulated reactor vessel which itself
is enveloped by a guard vessel. This guard vessel is a closed vessel that envelops the coreunit, providing containment and cooling through its vessel wall. Overheating of the core-unit
will cause the core-unit to heat up and increase heat loss from the core-unit, in turn increasingheat transfer, via thermal radiation, to the guard vessel. The guard vessel in turn is surrounded
by a robust air cooling jacket. This cooling jacket will provide long term cooling. The coolingjacket operates at atmospheric pressure so will continue to cool in the event of a leak or
damage to the jacket. This is illustrated in Figure 3.
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Reactivity Control
Reactor criticality control is assured through a negative temperature feedback made possible
by molten salt fuel. This negative temperature feedback assures reactor safety on overheating,even with loss of all control systems. Molten salt fuel does not degrade by heat or radiation,
so there are effectively no reactor power limits to the fuel.
For the control function, redundant, shutdown rods are also integrated into the IMSR coreunit. These shutdown rods will shut down the reactor upon loss of forced circulation and will
also insert upon loss of power. Another backup is provided in the form of meltable cans,filled with a liquid neutron absorbing material that will shut down the reactor on overheating.
5. Safety Features
The basic design approach to safety in the IMSR is to achieve an inherent, walk-away safe
nuclear plant. No operator action, electricity, or externally-powered mechanical components
are needed to assure the most basic safety functions:• Control: failure of control systems or reactivity insertion events only leads to reactor
stabilization at a slightly higher temperature.
• Cool: inherent heat sinks are available initially to absorb transient and decay heat,with heat losses providing long term cooling for both core and containment.
• Contain: the fluoride salts are chemically stable, bind radioactive fission products tothe salt, and have extremely high boiling points. Multiple engineered barriers areprovided as backup to this inherent containment. Partial entrenchment of the reactor,
combined with thick concrete and steel shielding, provides resistance to external
events such as earthquakes, explosions, and aircraft crash.
An important part of the IMSR safety philosophy is to start by removing drivers that pushradioactive material into the environment. Specifically the reactor always operates at low
pressure due to the inert, low-volatile fuel-coolant mixture and the absence of water or steam
in the reactor. This approach completely eliminates stored energy, both physical andchemical, from the reactor system. The IMSR further augments this high level of inherent,
physics-based safety with its integrated, pipe-less, fail-safe systems architecture.
The IMSR does not depend on depressurizing the reactor or bringing coolant to the reactor.
All required control and heat sink functions are already present where they are needed – inand directly around the IMSR core-unit. As such the IMSR completely eliminates any
dependence on support systems, valves, pumps, and operator actions. This is the case in boththe short term and the long term. To make this possible, the IMSR designers have combinedmolten salt reactor technology with integral reactor design and a unique cooling system.
Fission Product Retention
The main barrier for fission product release is the fluoride salt that is chemically stable andthat bind radioactive fission products to the salt. The core-unit represents the second barrier.
The guard vessel containment is provided as an additional hermetic barrier in the extremely
unlikely event that the integral core-unit itself would experience major failures. Without
sources of pressure in the core-unit or in the containment itself, the containment is neverchallenged by pressure. Overheating of the containment is precluded by the balance of heatlosses and heat generation even if the core-unit would fail. The containment itself is coveredby thick horizontal steel radiation shield plates at the top. These plates also provide protection
against extreme external events such as aircraft crash or explosion pressure waves, andprovide an additional heat sink in any overheating scenario.

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6. Instrumentation and Control Systems
The IMSR does not depend on electrical or even I&C systems for ultimate safety. For
operability and investment protection, a few additional requirements do exist, such as freeze
protection, as the fluoride salts have high freezing points. The normal plant operations
systems provide this protection, in the form of pump trips, electrical (trace) heating, andthermocouple sensors. As with the steam turbine-generator, this type of equipment isavailable off-the-shelf from many suppliers. The pumps are the primary control units, aspassive, flow driven control rods are utilized. This links the trip and protection logic to thepump rather than to a control rod drive and logic system. The result is a simple and easy tocontrol system where the only real variable to control actively is the pump speed (trace
heating is typically kept in automatic mode for normal operations). For example, to rapidly
shut down the reactor, no control rod drives, drive power, hydraulic supply, or other support
systems are needed; instead the pumps are simply tripped. The simplicity of this approachand the lack of dependence on any electrical systems also mean that cyber-security is easy toaddress.
As the IMSR does not depend on the electrical systems for ultimate safety, the design of adetailed plant electrical and control room system is eased. However, as with the steam
turbine-generator, standard, off-the-shelf industrial equipment is expected to be used due tothe generic nature of the requirements.
7. Plant Arrangement
The reactor auxiliary building and plant layout is shown in Figure 4. The IMSR core-unit ishoused in a below grade silo. Surrounding the core-unit are the guard vessel cooling jacket,
and at the top, a removable steel containment head plate. All these components are locatedbelow grade as well. Above the containment head, shield plates cover the core-unit forradiation shielding. An idle silo is provided next to the active silo, so that this silo can beloaded with a new core-unit while facilitating cool down of a spent core-unit. These two silos
make up the main components of the reactor building. After draining the spent core-unit
followed by a cool down period, the empty spent module is the lifted out and stored in longterm storage silos in an area adjacent to the reactor building. Inlets and outlets are provided
for the (non-radioactive) coolant salt lines coming into and out of the core-unit, whichtransfer the heat to steam generators and re-heaters. The coolant salt lines are located inside
the reactor building. The steam and reheat steam are utilized in a conventional, off-the-shelf
turbine-generator system located in an adjacent building.
8. Plant Performance
The simple modular and replaceable core-unit should improve reliability and plant
performance The core-unit has only a few moving parts (pumps) and these are redundantly
fitted. Even premature replacement of a faulty module only mildly affects reliability andlevelized cost. The use of an idling silo greatly increases plant capacity factor, as long cool
down times (for radiation dose reduction) are possible without plant downtime.
The IMSR reactor building has a low profile and low mass, allowing rapid construction. The
reactor building is a simple lightweight industrial building as it does not serve any major
safety-related function - these are all provided by dedicated, robust, fail-safe systems.
Being a molten salt reactor (MSR), the IMSR has low fuel reload costs; fuel fabrication costs
are zero, only extensive purification of the salts is required which, being bulk chemical
processing, has a much lower cost than traditional high precision fuel fabrication with all its

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quality and process control costs. Fuel recycling costs are also much lower with molten salt
fuels, as costly fuel deconstruction and – reconstruction steps are avoided and also because
simple, compact, low waste volume creating distillation and fluorination processes can beutilized. This makes it likely that spent IMSR fuel will be recycled.

Figure 4: Cut-away view into the reactor auxiliary building and plant layout(Reproduced courtesy of Terrestrial Energy)
9. Design and Licensing Status
The current phase of work involves the support of a growing number of universities, third
party laboratories and industrial partners. In addition, the Canadian nuclear regulator, theCanadian Nuclear Safety Commission, has started Phase 1 of a pre-licensing Vendor Design
Review of the IMSR. This is an assessment of a nuclear power plant design based on a
vendor's reactor technology.
[1] For more information: www.terrestrialenergy.com

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MSTW (Seaborg Technologies, Denmark)1. Introduction
The Seaborg Technologies’ Molten Salt Thermal Wasteburner (MSTW) is a thermal
spectrum, single salt, molten salt reactor, operated on a combination of spent nuclear fuel(SNF) and thorium. It is envisioned to produce 100 MWe, or 115 MWe with a 2 stage turbine,from 270 MWth. The core outlet temperature is 700°C, but can go as high as 900°C for
special uses, such as hydrogen production. The MSTW is designed around inherent safety
features; no active measures are required to control the reactor under abnormal
circumstances. The fully modularized MSTW is suitable for mass production. As a module
reaches the end of its lifecycle, it will be extracted and returned for recycling in a central
production facility after it has cooled down. The reactor core, including the graphite-based
moderator, is projected to have a lifecycle of seven years, while the power plant will operateon the same batch of SNF for the 60 year facility lifetime. The MSTW is in the early designphase and Seaborg Technologies (ST) is focused primarily on neutronics, radiative transfer,computational fluid dynamics (CFD), and the physics of the design. A model of theunderground reactor cave is shown in Figure 1.

Figure 1: Early model of the underground reactor cave
(Reproduced courtesy of Seaborg Technologies).
2. Target Application
The MSTW is designed for electricity production, district heating/cooling, sea water
desalination, etc. Due to the high outlet temperature it is well suited for synthetic fuel, as wellas industrial process heat applications. Its high burnup and the fact that it is fuelled directlywith spent nuclear fuel makes it a good option for spent nuclear fuel stockpile reduction.
However, it can, without modification, operate on a wide array of different fuels.
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3. Development Milestones
2014

Thermal wasteburner concept coined.

2015

Thermal wasteburner neutronic benchmark.
Pre-conceptual design (whitepaper).

4. General Design Description
Design Philosophy
The MSTW was originally designed to address the three main concerns that resulted in theDanish ban on nuclear power in 1985, namely safety, waste, and proliferation. As the design
has been developed and matured, the following three concepts together have been theessential philosophy around which the other features have been planned and implemented:
Safety: The system is placed in an optimised configuration, whereby any perturbation to sucha system - from operator mistakes to major earthquakes - will move the system away from theoptimum and thus result in an automatic power reduction and, if the situation is notalleviated, eventually reactor shutdown – see an illustration of one of these principles in
Figure 2.

Figure 2: Optimal configuration principle and safety through physics – illustrated through acore geometry optimization of keff (Reproduced courtesy of Seaborg Technologies).
Waste: By combining SNF with thorium, the production of transuranic elements (TRU) is
lower than the rate at which they are created. The net reduction is roughly five tonnes TRU
over the 60 year plant lifetime.
Non-proliferation: To ensure high proliferation-resistance the MSTW is designed to be asingle salt reactor wherein the chemical reprocessing system is incapable of extracting
specific actinides. Furthermore, no actinide element will have an isotopic composition usable
for nuclear weapons at any point in the fuel cycle.
Power Conversion Unit
The MSTW utilizes a coolant salt based on a core-integrated primary heat exchanger system.
The coolant salt acts as heat storage and flows through the primary side of the steam
generator, which delivers steam at 550°C to the single or two stage turbines (Rankine). The
coolant salt and heat exchanger system also serves as core shielding and neutron reflector.
The reactor core will passively shut down if overheated (negative thermal coefficient);
therefore, loss of coolant accidents (LOCAs) will result in reactor automatically powering
down. In case of a continued inability to cool the core, the fuel will eventually drain itself
through a freeze plug to a passively cooled dump tank.
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MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Seaborg Technologies

Country of origin

Denmark

Reactor type

Molten Salt Reactor

Electrical capacity (MWe)

100 (1-stage turbine) or 115 (2-stage turbine)

Thermal capacity (MWth)270

Expected capacity factor (%)

97.5%

Design life (years)

7 years component, 60 years plant life.

2

Plant footprint (m )20000

Coolant/moderator

Graphite basedPrimary circulation

Forced circulation

System pressure (MPa)

0.1 (sub-atmospheric)oCore inlet/exit temperatures ( C)

600/700 – or 700/900 in the high temperature configuration

Main reactivity control mechanism

Negative temperature coefficient; graphite control rods

RPV height (m)5

RPV diameter (m)4RPV or module weight (metric ton)

158 (including fuel and coolant)

Configuration of reactor coolant system

Integrated

Power conversion process

Rankine cycle

Process Heat Capabilities (High / Low T)

Different configurations possible (850°C /550°C)

Passive Safety Features:

Many: negative temperature coefficient; freeze plug to drain
fuel to passively cooled dump tank; overflow system, etc.

Active Safety Features:

Redundant, but, SCRAM and borofluoride injection installed.

Fuel salt

Sodium-actinide fluoride (93% Th, 3.5% U 3.5% Pu)

Moderator height (m)2.9

Number of moderator assemblies

91

Fuel enrichment (%)

Pre-processed SNF (U 1.1% fissile, Pu 69 % fissile)

Fuel burnup (GWd/ton)

250 (U and Pu - negligible Th cycle burning)

Fuel cycle (months)

720

Approach to engineered safety systems

Passive

Emergency Safety Systems

Inherent – passive

Residual Heat Removal System

Radiative (core), conduction (dump tank)

Refuelling outage (days)

Not Applicable(60 day module exchange, every 7 years)

Distinguishing features

Thermal spectrum wasteburner; Integrated heat-exchangers;
Power production controlled by coolant pump speed

Modules per plant

1 or more

Target construction duration (months)

36

Design Status

Conceptual (mainly neutronics benchmarks done)

Reactor Core
The MSTW core is a graphite-based compound-moderator thermal reactor core as illustrated
in Figure 3. The core is slightly over-moderated to ensure negative void and temperature
reactivity coefficients. The graphite is coated with a metal and corrosion is reduced by333regularly circulating the fluoride salt through a fluoride burner and by adding a reducing
anode to the fuel salt. The reactor operates at 12 kW per litre core, with a peak power densityof 250 kW per litre fuel near the core centre. Swelling and corrosion of the graphite
moderator is expected to be the limiting factor for the lifetime of the core module.

Figure 3: View of the inner reactor core and graphite moderator block (Reproduced courtesyof Seaborg Technologies).
Fuel Characteristics
The fuel salt is currently a eutectic sodium-actinide fluoride salt mixture, but sodiumrubidium-actinide or sodium-zirconium-actinide fluoride salts are being considered to reduce
the actinide fraction in the salt, which will decrease the power density in the fuel salt andreduce pumping power requirements. Initially the actinide fuel mixture is 93% thorium, and
7% pre-processed SNF (3.5% uranium (1.1% enrichment) and 3.5% reactor-grade
plutonium), but over time the plutonium “quality” decreases, and so only SNF is added andthe thorium fraction decrease.
Reactivity Control
The strongly negative temperature coefficient ensures a saturation temperature level of thecore, thus the power can be reduced by reducing cooling pump speed, or vice versa. The
online refuelling slowly feeds fresh SNF into the salt, for continuous operation. Daily
adjustment of saturation temperature level is achieved using a series of graphite fine-tuningrods, which can be inserted or extracted from the reactor. Using these rods, the reactor can be
maintained critical at its operational temperature for up to two months, without any additionof fresh SNF.
A SCRAM system exists for the operators’ convenience and for the 7-yearly core module
replacement scheduled for each seven years of operation.

334

Figure 4: Simulation of reactivity control using the chemical system. New spent fuel is added
while fission products are continuously extracted, handled and vitrified.
(Reproduced courtesy of Seaborg Technologies)
Reactor Pressure Vessel and Internals
The reactor operates below atmospheric pressure above the fluid surface. However, the core,
chemical unit, dump tank, and other critical components are planned to be contained in an
airtight dome, able to withstand a significant pressure, both from internal and external events.
5. Safety Features
In the MSTW safety is guaranteed by physics rather than engineering. The reactor relies onnatural convection, radiative heat transfer, and gravity for safety. The neutronics of thereactor is designed to be under optimal conditions. Then, any unexpected change to thesystem will result in a move away from the optimum, and thus a power reduction or
shutdown of the reactor. All active safety systems have been made redundant.
Emergency core cooling system and decay heat removal system
In case of LOCA the core will heat up and the strong negative temperature coefficient will
reduce reactivity and terminate power production without any operator interference. After
shut down the core will heat up from residual heating, and passive cooling will be ensured
through natural convection and radiative heat transfer for a substantial time. However, if main
cooling is not restored, the reactor fuel will eventually drain itself (actively or passively) bymelting a salt freeze plug. In the drain tank the fuel will solidify and is cooled by conduction
to an underground heat sink.
Containment system
With the exception of noble gasses, noble metals, and cadmium, fission products are nonvolatile in a fluoride salt. Note that this includes the hazardous caesium and iodine. The
MSTW is designed so that critical temperatures are physically unreachable due to a numberof reasons including the negative temperature coefficient, and the balancing of residual
heating and radiative heat transfer. As such there are no scenarios where large amounts of
non-volatile fission products can escape the fluid – even if the entire containment is breached.
335

The remaining volatile fission products are continuously extracted, handled and vitrified, andare therefore never present in significant amounts in the fluid. For this reason, the source
term, even during extreme disasters, will be small and will pose a minimal off-site risk.
Despite the inherent safety, the reactor is designed with three barriers, where no irradiatedcomponent, coolant or fuel leaves the outer two barriers at any point.
6. Plant safety and Operational Performances
The neutronic optimum design philosophy provides excellent transient behaviour andoutstanding inherent safety features, and the plant has outstanding load following capabilities.
The heat storage in the coolant salt and the automatic saturation temperature of the reactor
enables power production to be turned on and off as fast as the steam generators can beswitched on and off. Furthermore, due to the degassing of noble gasses, and the relatively
large fuel inventory, the reactor will not experience a poison-out during shutdown or power
reduction.
7. Instrumentation and Control systems
The MSTW will have a series of control and monitoring features. Control will be based on aseries of fine-tuning moderation rods, which will be placed near the edge of the reactor coreso that an ejection event will reduce reactivity (loss of reflector). The reactivity span of theserods will be such that the reactor core saturation temperature can be adjusted within 200°C,
and such that critical temperatures are not reachable through intentional or unintentional
operator failures.
8. Plant Arrangement
Plant design is not yet finalized. However, all irradiated components, fuel, and coolant salt
are contained in an approximately 8 m deep and 12 m wide underground area, covered with 2
m of concrete during operation. The turbine building, control room, and module handlingarea are placed above ground, above and around the underground area. The plant isenvisioned to be operated by two operators, plus maintenance personnel for turbines andother plant components. During module change out, a team from the central manufacturingsite will be dispatched to the site.
9. Design and Licensing Status
ST is currently focused on advanced multi-physics, neutronics, and early engineering details.
The facility is planned as a green-field concept, and decommissioning considerations feature
heavily in the design process, and most component parts are recyclable. Licensing is
considered, and a plan has been drafted, however, at this early stage it is of lesser concern.
10. Plant Economics
The design is at a conceptual stage and the details of the plant economics are yet to beassessed. It is our belief that the redundancy of all active systems will result in significant
cost reductions, through fewer engineering requirements and general simplicity. Also, thecentralized manufacturing and recycling of modules is expected to benefit plant economics.
Lastly, the short lifecycle of modules will heavily reduce requirements of materials andengineering. This is expected to significantly reduce the overnight cost, but will come at theexpense of a larger over time cost. Postponing a large fraction of the plant cost from theovernight cost will be favourable from an economical point-of-view.

336ThorCon (Martingale, International
Consortium)
1. Introduction
ThorCon is a molten salt nuclear reactor [1]. Unlike all current operating reactors, the fuel isin liquid form. The molten salt can be circulated with a pump, and passively drained in theevent of an accident. ThorCon operates at slightly above atmospheric pressure (similar to thatin one’s garden hose) and can thus use normal pipe thicknesses and easily automated, shipstyle steel plate construction methods. The entire nuclear part of the plant is underground, as
shown in Figure 1. This drawing shows a 1GWe ThorCon composed of four 250 MWe
modules. The decay heat cooling towers are on the left. The underground nuclear island iscenter left. The yellow rectangles are hatches and are served by gantry cranes. The turbogenerator halls are center, and the switchyard is on the far right. The main cooling towers, if
required, are to the right of the switchyard. The cranes allow periodic replacement of all
critical components including the reactors and fuelsalt. The reactors and fuelsalt aretransported by a special-purpose ship shown in the background.

Figure 1: Birdseye View of 1 GWe ThorCon (Reproduced courtesy of ThorCon)
2. Target Application
The first planned application of ThorCon reactors is to produce electricity in developing
nations. As such, the presumption is that the grid is still developing and the reactor must be
capable of off-grid startup and thus must not depend on grid power. Cost is critical in thesemarkets: ThorCon must be cost competitive with coal and be deployable at least as rapidly. A
pre-feasibility study was initiated in Indonesia in 2016.
3. Development Milestones
2011

Conceptual design development

2016

Pre-feasibility study in Indonesia

2018

Pre-fission testing starts

2020

Fission testing starts

2022

Commercial operation starts

337

4. General Design Description
Design Philosophy
The following principles are followed in the ThorCon design [2,3]:
ThorCon is Walkaway Safe
ThorCon is a simple molten salt reactor with the fuel in liquid form. If the reactor overheats
for whatever reason, it will automatically shut itself down and drain the fuel from the primaryloop and passively remove the decay heat. There is no need for any operator intervention andthe operators cannot prevent the draining and cooling. The reactor is 15 m underground.
ThorCon has three gas tight barriers between the fuelsalt and the atmosphere. The reactor
operates at slight over-pressure so that in the event of a primary loop rupture, there is nodispersal energy and also no phase change. The spilled fuel merely flows to a drain tankwhere it is passively cooled. The most troublesome fission products, including 90Sr and 137Cs,
are chemically bound to the salt. They will end up in the drain tank as well.
ThorCon is Ready to Go
The ThorCon design should not need new technology development. ThorCon is a scale-up of
the successful Molten Salt Reactor Experiment (MSRE). Currently the designers foresee notechnical reason why a full-scale 250 MWe prototype cannot be operating within four years.
The intention is to subject this prototype to all the failures and problems that the designers
claim the plant can handle. As soon as the prototype passes these tests, commercial
production can begin.
ThorCon is Rapidly Deployable
The entire ThorCon plant including the building is designed to be manufactured in blocks ona shipyard-like assembly line. These 150 to 500 ton, fully outfitted, pre-tested blocks are thenbarged to the site. A 1 GWe power station will require less than 200 blocks. Site work islimited to excavation and erecting the blocks. This should result in order of magnitudeimprovements in productivity, quality control, and build time. A single large reactor yard canturn out one hundred 1 GWe ThorCons per year. The philosophy is therefore that ThorCon ismuch more than a power plant; it is a new system for building power plants.
ThorCon is Fixable
The design does not foresee any complex repairs to be attempted on site. Except for thebuilding everything else in the nuclear island is replaceable with little or no interruption in
power output. Every four years the entire primary loop is changed out, returned to acentralized recycling facility, decontaminated, disassembled, inspected, and refurbished. The
instrumentation design and monitoring system is designed to identify incipient problems
before they can lead to failures. Major upgrades must be possible without significantly
disrupting power generation. A nuclear power plant following such a change-out strategy canin principle operate indefinitely; but decommissioning should be little more than removing,
but in this case not replacing, all the replaceable parts.
ThorCon is Cheaper than Coal
ThorCon requires far less resources than a coal plant. Assuming efficient, evidence-based
regulation, ThorCon aims to produce clean, reliable, carbon-free electricity at less than thecost of coal.
Nuclear Steam Supply System
Figure 2 shows a cutaway view of the underground structure. ThorCon is divided into
250 MWe power modules. The drawing shows two such modules. Each module contains two338

replaceable reactors in sealed Cans. The Cans are depicted in red in the drawing. They sit insilos. At any one time, just one of the Cans of each module is producing power. The otherCan is in cooldown mode. Every four years the Can that has been cooling is removed andreplaced with a new Can. The fuelsalt is transferred to the new Can, and the Can that has
been operating goes into cool down mode.

Figure 2: Cutaway view of two module silo hall (Reproduced courtesy of ThorCon)
Figure 3 takes a look inside a Can. The Can
contains the reactor, which we call the Pot, aprimary loop heat exchanger (PHX), and a
primary loop pump (PLP). The pump (blueupper left) takes liquid fuelsalt — a mixtureof sodium, beryllium, uranium and thorium
fluorides — from the Pot (orange) at 704oC,
and pushes the fuelsalt over to the PHX at a
rate of just under 3000 kg/s (1 m3/sec).
Flowing downward through the PHX(skinny blue), the fuelsalt transfers heat to asecondary salt, and is cooled to 564 oC in theprocess. The fuelsalt then flows over to thebottom of the Pot, and rises through thereactor core where the graphite moderator
slows the neutrons produced by the fissile
uranium, allowing a portion of the uranium
in the fuelsalt to fission as it rises throughthe Pot, heating the salt and (indirectly)
converting a portion of the thorium to fissile
uranium.

Figure 3: The ThorCon can: a pot, a pump,
and a still (right)(Reproduced courtesy of ThorCon)

339

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Martingale

Country of origin

International consortium planning first deployment in Indonesia

Reactor type

Thermal Molten Salt Reactor

Electrical capacity (MWe)250 (per module)

Thermal capacity (MWth)

557 (per module)

Expected capacity factor (%)> 90%

Design life (years)

80 years

2

Plant footprint (m )20000 for 500 MWe

Coolant/moderator

NaF, BeF2 salt, graphite moderated

Primary circulation

Forced circulation

System pressure (MPa)0.3 at primary loop max stress point, 1.05 at exit of primary

Core inlet/exit temperatures (oC)565 / 704

Main reactivity control mechanism

Negative temperature coeff; salt flow rate, control rod insertion

RPV height (m)/ diameter (m)12 (includes full primary loop and off-gas) / 8

RPV or module weight (metric ton)

400

Configuration of reactor coolant system

Four loops: Fuel salt, secondary salt, solar salt, steam.

Power conversion process

Rankine steam

Passive Safety Features:

Fully passive shutdown and cooling. 72 day grace period.

Active Safety Features:

Drain fuel salt, shutdown rods.

Fuel salt

12% heavy metal in NaBe salt.

Heavy metal composition

80% Th, 16% U-238, 4% U-235

Makeup salt

12% uranium (enriched to 19.7%) in NaBe salt

Fuel enrichment (%)

19.7

Fuel burnup (GWd/ton)

256 GWd/ton U

Fuel cycle (months)

96

Approach to engineered safety systems

Avoid them. Physical limit on fuel addition rate; H/W limit onpump speed change rate.

Number of safety trains

Three means to remove decay heat. Two are fully passive.

Emergency Safety Systems

Three levels of containment, 3 cooling systems, 2 shutdown

Residual Heat Removal System

primary cooling to ocean; natural circulation to air; steam release

Refueling outage (days)

approximately 7

Distinguishing features

Low cost, full passive safety, short construction time

Modules per plant

1-4 per building, arbitrary per generating station

Target construction duration (months)

6

Seismic design

Target 0.8 pga

Design Status

Finishing conceptual design

340

The Pot pressure is 3 bar gage at the maximum stress point. The outlet temperature of 704 oC
results in an overall plant efficiency of about 45%, and a net electrical output per Can of
250MW. The Can’s net consumption of fissile uranium is 112 kg per year. The Can (red) is acylinder 11.6 m high and 7.3 m in diameter. It weighs about 400 tons. The Can has only one
major moving part, the pump impeller.
Directly below the Can is the Fuelsalt Drain Tank (FDT) (green). In the bottom of the Can isa fuse valve (grey). The fuse valve is merely a low point in a drain line. At normal operating
temperatures, the fuelsalt in the fuse valve is frozen creating a plug. But if the Can heats up
for any reason, the plug will thaw, and the fuel salt will drain to the FDT. Since the drain tankhas no moderator, fission will stop almost immediately. This drain is totally passive. There isnothing an operator can do to prevent it.
A critically important feature of ThorCon is the silo cooling wall (blue). The silo wall ismade up of two concentric steel cylinders, shown in blue in Figure 4. The annulus between
these two cylinders is filled with water. The top of this annulus is connected to a condenser ina decay heat pond. The outlet of this condenser is connected to the basement in which the
Can silos are located. This basement is flooded. Openings in the bottom of the outer silo wall
allow the basement water into the bottom of the annulus. The Can is cooled by thermalradiation to the silo cooling wall. This heat converts a portion of the water in the wall annulus
to steam. This steam/water mixture rises by natural circulation to the cooling pond, where thesteam is condensed, and returned to the bottom of the cooling wall via the basement. In thisprocess, some of the water in the pond is evaporated. The decay heat cooling towers returnalmost all this water to the pond.

Figure 4: Silo Hall Cross-Section (Reproduced courtesy of ThorCon)

341

The silo cooling wall also cools the Fuelsalt Drain Tank (FDT). The drain tank is tall, thinrectangular trough that has been wrapped into a circle. This arrangement provides sufficientradiating area to keep the peak tank temperature after a drain within the limits of the tankmaterial. This cooling process is totally passive, requiring neither operator intervention nor
any outside power.
Each Can is located in a Silo. The top of the Silo is 14 m underground. Figure 4 shows thesecondary salt loop in green. The secondary salt is a mixture of sodium and beryllium
fluoride containing no uranium or thorium. Hot secondary salt is pumped out of the top of thePrimary Heat Exchanger to a Secondary Heat Exchanger where it transfers its heat to a
mixture of sodium and potassium nitrate commonly called solar salt from its use as an energy
storage medium in solar plants. The solar salt, shown in purple in Figure 4, in turn transfers
its heat to a supercritical steam loop, shown in red.
ThorCon is a high temperature reactor that translates to thermal efficiency of up to 45%
compared to about 32% for standard light water reactors. This reduces capital costs and cutscooling water requirements by 60%. It also allows us to use the same steam cycle as amodern coal plant.
Reactor Core
The reactor core is inside the pot as shown in Figure 3. The core is 90% filled with graphite
slabs, the moderator. The core is 5m diameter, and 5.7m high.
Fuel Characteristics
The fuelsalt is NaF-BeF2-ThF4-UF4 76/12/9.5/2.5 where the uranium is 19.7% enriched. As
fissile is consumed more fissile (either U233 or Pu239) is generated but not enough to replace
the fuel burned. The reactor has no excess reactivity, no burnable poisons, and no poison
control rods. Makeup fuel must be added daily.
Fuel Handling System
Makeup fuel is added by applying gas pressure to the makeup fuelsalt tank which forces
makeup fuel into the primary loop. The makeup fuel composition is NaF-BeF2-UF4 76/12/12
where the uranium is 19.7% enriched. The makeup fuel addition rate is physically limited toensure the adding reactivity rate stays within acceptable limits. Excess fuelsalt flows into aholding tank.
Reactivity Control
The primary reactivity control is temperature and fuelsalt flow rate. For slow reactivity
control makeup fuelsalt (or makeup fertile salt with no fissile) additions allow modest daily
increase or decrease of the reactivity.
Reactor Pressure Vessel
The ThorCon reactor is never under high pressure so that the typical term Reactor Pressure
Vessel does not really apply. In the design the Can plus the Fuel Drain Tank fulfil the samefunction since all radioactive material (except tritium) should be contained within thesestructures. Since no high pressure is present that can act as a driving force to spread thecontent into the environment, the RPV does not have the central safety importance in an MSR
that it does in a LWR.

342

5. Safety Features
The ThorCon design combines a negative temperature coefficient with a large margin
between the operating temperature of 700 oC and the fuelsalt’s boiling temperature (1430 oC)
to assure passive, totally unavoidable, shutdown and cooling. In any event that raises thetemperature of the salt much above the operating level, the reactor will automatically shutitself down. If the high temperature persists, the fuse valve will thaw and drain the fuel fromthe primary loop to the drain tank, where the silo cooling wall will passively remove thedecay heat.
There is no requirement for operator (or control system) intervention at any time since there
are no valves to realign, pumps to activate, or any other actions to be taken. In fact there isnothing the operators can do to prevent the drain and cooling. The decay heat is transferred toan external pond which has sufficient water for 72 days cooling. After 72 days without any
intervention the water in the pond will be running low. Adding more water is simple since thepond is accessible and at atmospheric pressure. If the pond cooling line is lost, there isenough water in the basement to handle the first 30 days of decay heat
Release Resistant The ThorCon reactor is 15 m underground. ThorCon has three gas tightbarriers between the fuelsalt and the atmosphere. Two of those barriers are more than 10 m
underground. ThorCon reactor operates at near-ambient pressure. In the event of a primaryloop rupture, there is no dispersal energy and no phase change and no vigorous chemical
reactions (like zirconium and steam). The spilled fuel merely flows to the drain tank where itis passively cooled. Moreover, the most troublesome fission products, including iodine-131,
strontium-90 and cesium-137, are chemically bound to the salt. They will end up in the draintank as well. Even if all three barriers are somehow breached, almost all these salt seekers
will not disperse.
No separate, spent fuel storage ThorCon uses an eight-year fuelsalt processing cycle, after
which the used salt is allowed to cool down in the non-operating Can for four years,
eliminating the need for a separate, vulnerable spent fuel storage facility. The fuelsalt that iscooling is as well-protected as the fuelsalt that is currently being burned.
Four loop separation of steam and fuelsalt. ThorCon employs four loops in transferring heatfrom the reactor to the steam turbine. The solar salt loop captures any tritium that has made it
to the secondary loop, and more importantly ensures that a rupture in the steam generator
creates no harmful chemicals and harmlessly vents to the Steam Generating Cell via an open
standpipe.
6. Plant safety and Operational Performances
Load following is accomplished by changing the pump speed while keeping the temperatures
relatively constant. The requirements for load following are drawn from EU Utility
Requirements for Light Water Reactors (5% full power per minute from 50% to 100% full power).
Since the off-gases are continuously removed xenon poisoning and oscillations are not an
issue. No neutron poisons are used in the control of the reactor which reduces fuelconsumption.
7. Instrumentation and Control systems
The instrumentation and control and even the operators are not safety critical in the design.
ThorCon instrumentation uses a multiplicity of sensors to record and report the condition of
the power plant. Statistical process control is used to track trend lines and detect incipient
failures such as bearing wear. A common control center for all modules in a power plant343

minimizes staffing requirements. Continuous reporting of plant conditions to a central
engineering facility monitoring conditions at all plants allows for fleet wide analysis andmakes expert advice available to any plant experiencing unusual conditions as well as
monitoring plants for unusual activity.
8. Plant Arrangement
Sections 1 and 4 and Figures 1-4 describe the plant arrangement. The control building isshared among all modules in a power plant. In the most common deployment envisaged two
power modules will drive a single 600 MWe turbine/generator. This size was chosen as itshould be competitive priced per MWe while also suitable for smaller deployments.
9. Design and Licensing Status
Conceptual design is nearly complete. License discussions have been started with Bapeten(Indonesian regulator).
10. Plant Economics
In order to be successful, ThorCon must be able to compete with coal with zero CO2 price. A
first comparison was therefore made with the cost of coal plants. Since ThorCon uses thesame turbine island and electrical side as a standard 500 MWe supercritical coal plant, thecost comparison was made on the steam generation parts. When the steel and concrete
requirements of a 1 GWe plant was compared, the ThorCon nuclear island requires one-fifth
as much steel and one-third as much concrete as the coal plant boiler, fuel handling, and ash
handling systems.
Moreover, almost all the ThorCon concrete is non-structural, and simple construction of
concrete dumped into the steel sandwich walls; while a large part of the coal plant concrete isslow, labour intensive, reinforced concrete. Based only on this resource basis the overnight
cost of the ThorCon nuclear island should be less than one-third that of the coal plant’s steam
generation side.
Other costs also need to be taken into account. A 1GW ThorCon requires 1500 tons of very
high quality graphite, 1300 tons of SUS 316, 220 tons of the superalloy Haynes 230, 2,500
tons of lead, and 200000 m3 of excavation. But these and other adjustments add a little morethan 100 million dollars to the cost of a 1 GWe plant or about $100/kW. Overall, the resource
cost of the ThorCon nuclear island is less than one-half that of the coal plant’s steam
generation systems, or less than a $500 per kW. Based on this assessment ThorCon is cheaper
than coal, both in capital costs and in fuel costs.

[1] Complete ThorCon website: http://thorconpower.com[2] Succinct description of philosophy and design: http://thorconpower.com/docs/domsr.pdf[3] Designers’ essays on molten salt reactor aspects:
http://thorconpower.com/library/documents

344

FUJI (International Thorium Molten-Salt
Forum, Japan)
1. Introduction
The Molten Salt Reactor (MSR) uses molten salt, in general molten fluoride salt, as liquid
fuel and coolant. 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 (now, International Thorium
Molten-Salt Forum: ITMSF), based on the ORNL’s results with the view to deploy it widelyin the world [1,2].
Molten salt is a liquid, which is in general a melted chemical compound of acid and alkali athigh temperature. 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 100 MWe to 1000 MWe. But, a latest and typical design(FUJI-U3) is 200 MWe, which can be categorized as small-sized reactors with modular
designs (SMR). The thermal output of FUJI-U3 is 450 MWt and thus a 44% thermal
efficiency can be attained. In addition to the expected high efficiency the simple core
structure and high fuel efficiency should facilitate a favourable economic performance.
A schematic diagram of the MSR-FUJI is shown in Figure 1.

Figure 1: Schematic diagram of MSR-FUJI (Reproduced courtesy of ITMSF)

345

Molten fuel salt can contain thorium (Th) as fertile material and 233U as fissile material, andbased on this the FUJI-U3 design can attain a self-sustaining fuel cycle with a conversion
factor of 1.0. Since MSR-FUJI applies the Th-cycle, generation of plutonium (Pu) and minor
actinide (MA) is very small compared with Light Water Reactors (LWR). Furthermore, it canconsume Pu as the fissile material, 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 shorterones.
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 bydesalination of seawater or for hydrogen production, utilizing its high exit temperature of
704 °C.
3. Development Milestones
1980’sConceptual designs of MSR-FUJI have been started

1980’s

Accelerator Molten-Salt Breeder (AMSB) design for a large production of fissile
material (similar to Accelerator Driven System ADS)

Until 2008

Several designs, such as a pilot plant (mini-FUJI), a large-sized plant (super-FUJI), a Pufueled plant (FUJI-Pu),

Recent

The latest SMR plant (FUJI-U3),
The MSR-FUJI as in this text refers to the 200 MWe FUJI-U3 design

4. General Design Description
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 andfurther 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 supercriticalpower 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 thickreactor vessel and pipes are not required. There are no fuel assemblies or complex coreinternal structure with the only component within a reactor vessel being the graphite
moderator. Based on these design principles in-factory fabrication would be simple.
Nuclear Steam Supply System
The nuclear steam supply system (NSSS) consists of a reactor core, pipes, pumps, a heatexchanger (HX), and a steam generator (SG), which supplies steam to a turbine/generator(T/G), as is shown in Figure 1. This figure shows only one loop, but a loop can be redundantdepending on a plant size or a need for flexibility.

346

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

International Thorium Molten-Salt Forum: ITMSFCountry of origin

Japan

Reactor type

Molten Salt Reactor

Electrical capacity (MWe)200

Thermal capacity (MWth)450

Expected capacity factor (%)

75% for daily load-following operation

Design life (years)

30

2Plant footprint (m )<5000 (R.B. + SG.B. + TG.B.)

Coolant/moderator

Molten fluoride / Graphite

Primary circulation

Forced circulation by pump

System pressure (MPa)

0.5 (by pump head)oCore inlet/exit temperatures ( C)

565 / 704

Main reactivity control mechanism

Control rod, or pump speed, or fuel concentration

Reactor Vessel height (m)5.4 (inner)

RV diameter (m)5.34 (inner)

RV weight (metric ton)

60 (made of Hastelloy N)

Configuration of reactor coolant system

Primary loop and secondary loop

Power conversion process

Rankine Cycle

Cogeneration / Process Heat Capabilities (High / Low T)

Possible, utilizing high exit temperature

Passive Safety Features:Freeze valve to drain tank.

Active Safety Features:

Control rod scram. ECCS is not required

Fuel type/assembly array

Molten salt with thorium and uranium

Fuel assembly active length (m)No fuel assembly

Number of fuel assemblies

No fuel assembly

Fuel enrichment (%)

2.0 equivalent (0.24%233U + 12.0%Th). Pu or LEU can be used.

Fuel burnup (GWd/ton)

No mechanical limitation for burnup

Fuel cycle (months)

Continuous operation is possible

Approach to engineered safety systems

Passive safety

Number of safety trains

ECCS/CCS/ADS are not required

Emergency Safety Systems

Ditto

Residual Heat Removal System

Passive cooling without electricity (required at drain tank)

Refueling outage (days)<30 (Refueling shutdown is not required)

Distinguishing features
Modules per plant

High safety, high economic performance, contribution to nonproliferation, and fuel cycle flexibility(Not decided)

Target construction duration (months)(Not decided)

Seismic design

Same as LWRs

Core damage frequency (per reactor-year)

Core meltdown is impossible

Design Status

3 experimental MSRs were built. Detailed design has not started

347

MSR-FUJI is designed to produce an exit temperature of 704°C in molten fuel salt, and its
heat is transferred to the secondary salt through a HX. Then, its heat produces 538°C
supercritical steam at a SG, and generates electricity by a supercritical T/G. Owing to its hightemperature, MSR-FUJI can achieve 44% thermal efficiency.
Since molten salt is used as fuel and coolant, loss of coolant accident (LOCA) may need a
new definition for MSR. In case of pipe break, leaked molten salt is drained to an emergency
drain tank without passing through a freeze valve. Even in this case, a molten salt loop
returns to atmospheric pressure when a pump stops, and a pressurization accident isincredible, 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.
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 inemergency conditions.
Reactor CoreA reactor vessel is cylindrical in shape. The core structure is made up by hexagonal shaped
graphite moderator blocks. The blocks contain holes that serve as the flow paths of themolten 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 fuelconcentration adjustment system. Since there are no fuel assemblies in the core, refueling
shutdown is not required, and continuous operation is possible. In order to achieve a core
conversion factor of 1.0 it is recommended to refresh the fuel salt every 7 years. Periodic
maintenance shutdown will of course be required as in any power plant.
Fuel Characteristics
The molten fuel salt is a liquid form of fluoride (LiF-BeF2) with ThF4 and a small amount of
233
UF4. A typical composition 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 temperatureand 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. Owing to unique features of molten fuel salt, fuel assembly
fabrication is not required, and radiation damage or fuel cladding failure does not occur.
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 bycore flow or by core temperature. Control rods are withdrawn in normal operation and areinserted 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 and the reactor vessel wall thickness is about 5 cm. Only
one component, the graphite moderator blocks, is present within the core internal region.

348

Secondary salt loop
The secondary loop adopts molten salt of NaBF4-NaF. This secondary loop is circulated by acentrifugal pump and removes heat from the primary loop through a heat exchanger to thesteam generator. Since the pressure of both the primary and secondary loop is very low, thedanger of rupture is minimized. In case of a pipe break molten salt is drained to a drain tank.
Steam Generator and Turbine Generator
A steam generator (SG) of MSR-FUJI adopts a shell and tube design. A U-shaped shell
contains a secondary salt flow, and steam flows inside of multiple tubes within a shell. The
secondary salt loop at 633°C provides heat to the SG that generates 252 kg/cm2 steam at
538°C fed to the supercritical turbine generator (T/G) to produces 200 MW electricity.
5. Safety Features
The MSR-FUJI design has very favourable safety characteristics that essentially exclude thepossibility of severe accidents based on the following unique features:• The primary and secondary loops are operated at a very low pressure, which
essentially eliminates accidents such as system rupture due to high pressure.
• The molten salt is chemically inert, that is, it does not react violently with air or water,
and is also not flammable. The corrosion of Hastelloy N can be minimized by theappropriate chemical control and maintenance of the molten salt.
• Pressure increase in a primary loop is incredible, because a boiling temperature of thefuel salt is very high (about 1,400°C) compared with the operating temperature (about
700°C).
• Since there is no water within the containment, there is no possibility of high pressure
by steam generation, and therefore no possibility of hydrogen explosions at any
accidental conditions.
• The fuel salt is drained to a drain tank through a freeze valve, if required. In case of arupture in the primary loop, the spilled fuel salt is drained to an emergency drain tankwithout passing a freeze valve.
• The fuel salt only reaches criticality within the graphite core (with the appropriate
moderation). In an accident the fuel salt is drained to a drain tank designed so that a
re-criticality accident does not occur.
• MSR has a large negative reactivity coefficient of a fuel salt temperature that can
suppress an abnormal change of the reactor power. Although a temperature coefficientof graphite is positive, it does not affect the safety, because the heat capacity ofgraphite is large enough.
• Since gaseous fission products (FP) can be removed by separation from the fuel salt,
the danger due to the release of radioactivity from the core at accidental conditions
can be minimized.
• Since fuel composition can be adjusted when necessary, the excess reactivity and thereactivity margin to be compensated by control rods are small. Therefore, reactivity
requirements for control rods are also small.
• The delayed neutron fraction of 233U is lower than that of 235U and some of delayed
neutrons are generated outside the core. However, safe control of the reactor ispossible because of a large negative reactivity coefficient and limited potential
reactivity insertion.
• Since there is no airflow and no heat source within the core when the fuel salt isdrained at accidental conditions, self-sustained graphite oxidation does not occur.

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Engineered Safety System Approach and Configuration
Many of the safety features were described above. Therefore, redundant and diverse
emergency core cooling systems (ECCS), makeup water pools, and automatic
depressurization system (ADS) are not required. In order to protect against a freeze accident
in a molten fuel salt loop, a high temperature containment is equipped.
Decay heat removal system
In normal shutdown condition, decay heat is transferred to a secondary loop and a steam-line
loop, and disposed to the ultimate heat sink (seawater for example). If all pumps in a primaryor secondary loop stop, fuel salt is drained to a drain tank through a freeze valve. Decay heatat the drain tank is cooled by a passive heat removal system, and finally its heat is disposed tothe outside environment through an air-cooled system that does not require electricity.
Emergency core cooling system
As is explained above, redundant and diverse emergency core cooling systems (ECCS) andmakeup water pools are not required. This would simplify the plant, and eliminate concernsof failures in safety systems.
Containment system
Since the risk of pressurization accidents is incredible, 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 MSRFUJI design has three levels of containment. The first is the reactor vessel and pipes made of
Hastelloy N. The second is a high temperature containment composed of three layers, whichcontains a reactor vessel, pipes, and a heat exchanger, as is shown in Figure 2. In order toavoid a freezing accident, this containment is equipped with heaters. The third level is areactor building composed of two layers. As explained above, a pressurization accident isincredible due the low vapor pressure. Therefore, a containment cooling system (CCS) andmakeup water pools are not required.

Figure 2: Vertical cross section of primary system of MSR-FUJI(Reproduced courtesy of ITMSF).
350

6. Plant safety and Operational Performances
Overall safety is described above. In case of a station blackout (SBO: Loss of all AC
electricity) the MSR-FUJI can be shut down and cooled without electricity. Core meltdown or
steam/hydrogen explosion is physically excluded by design and no ECCS is needed. As forlong-time operation, reactivity can be controlled anytime by a fuel concentration adjustment
system.
In normal daily operation, the power level can be controlled by core flow or by coretemperature control. That is, load following is easily performed without using control rods.
Control rods are withdrawn in normal operation, and are inserted by gravity in case of
emergency shutdown.
7. Instrumentation and Control systems
Instrumentation and control (I&C) systems in the MSR-FUJI design are the same as forrecent LWR designs. It must support operators in making decisions and efficiently operating
the plant during plant start-up, shutdown, normal operation, surveillance testing, andaccidental situations. It adopts the man-machine interface more useful, and expands the scopeof automatic control.
I&C systems include an alarm system, an information processing system, a reactor protection
system, an engineered safety equipment control system, power and process control systems,
sensors and instrumentation, an radiation monitoring system, and so on.
8. Plant Arrangement
Major buildings of MSR-FUJI are a reactor building, a steam generator building with a main
control room, and a turbine-generator building as shown in Figure 3. An auxiliary building is
not shown.

Figure 3: Bird-eye view of MSR-FUJI (Reproduced courtesy of ITMSF).
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 cylindrical351

shape with a hemispherical dome, which is made of concrete with steel liner as its inner layer.
The reactor building is founded on a common base-mat together with other buildings.
Control building
The main control room (MCR) is located at a steam generator building, which is next to areactor building. The MCR is a key facility to cope with normal and emergency situations, so
it is designed to ensure that plant personnel successfully perform the tasks according to theproper procedures.
Balance of plant:
Turbine Generator Building
The turbine generator (T/G) building contains the supercritical turbine and generator,
which produce electricity. Also, it contains condensers for disposed steam. The
condensers use outside water (for example, sea water) for cooling.
Electric Power Systems
These systems include the main generator, transformers, emergency diesel generators(EDG), and batteries. MSR-FUJI is equipped with two external electric sources for
operation, and EDGs are required for backup. In case of station blackout (SBO: Loss of
all AC electricity), it can be shut down and cooled without electricity.
9. Design and Licensing Status
Preliminary designs for various applications have been completed [ 3]. Three experimental
MSRs were constructed, and one of them was operated for 4 years without severe problems.
The detailed design has not yet started. Safety criteria and guidelines for MSR licensing areproposed [ 4].
10. Plant Economics
MSR-FUJI can achieve higher thermal efficiency owing to its high exit temperature. It has
the potential to be operated for longer periods with a high availability factor since it does not
require refueling shutdown. Also, fuel cycle advantages and cost saving can be realized due
to its self-sustaining fuel cycle and no need for fuel element manufacturing. A plant
construction cost for the 1000 MWe MSR-FUJI (super-FUJI) is estimated as less than$2000/KWe and the total electricity generation cost about 3 cents/KWh. In general the
economic performance of a smaller plant may be expected to be worse than that of a larger
plant. This factor has not been evaluated till now.
[1] Furukawa, K. et al. “Molten Salt Reactor for Sustainable Nuclear Power – MSR FUJI”.
p.821-856 of IAEA-TECDOC-1536 “Status of Small Reactor Designs Without On-Site
Refueling”, 2007
[2] Yoshioka, R. “Nuclear Energy Based on Thorium Molten Salt”, Chapter-23 of the book
“Molten Salts Chemistry: From Lab to Applications“ edited by F. Lantelme and H.
Groult, Elsevier Inc., USA, 2013[3] Yoshioka, R., Kinoshita, M. “Liquid Fuel, Thermal Neutron Spectrum Reactors”,
Chapter-11 of the book “Molten Salt Reactor”, Elsevier Inc., USA, to be published in
2017[4] Yoshioka, R., Mitachi, K., Shimazu, Y., Kinoshita, M. “Safety Criteria and Guidelines for
MSR Accident Analysis”, PHYSOR-2014, Kyoto, Japan, 2014352

Stable Salt Reactor (Moltex Energy,
United Kingdom)
1. Introduction
The Stable Salt Reactor (SSR) is unique in its use of molten salt fuel replacing solid pellets inconventional fuel assemblies. This brings the major advantages of safe molten salts without
the technical hurdles of managing a mobile liquid fuel. This combination should lead to
substantially lower capital costs than conventional nuclear power technologies. The SSR is a
design platform with two configurations, a thermal spectrum Low Enriched Uranium burner(SSR-U) and a fast spectrum actinide transmuting Wasteburning SSR (SSR-W). Both of these
designs can be adapted in the future to breed fuel from thorium. In the thermal versions, themoderator is graphite and forms an integral part of the fuel assembly thus eliminating
graphite lifetime concerns. The technology is coupled with large energy storage reserves toincrease peak power demand allowing an increase of renewables on the grid. This summary
focuses on the details of the thermal spectrum, uranium fuelled reactor SSR-U.

Figure 1: Isometric view of core module and section through reactor(Reproduced courtesy of Moltex)
2. Target Application
Stable Salt Reactor technology is initially aimed at mature markets with nuclear expertise.
Once deployed and proven to be commercially viable the large market potential is indeveloping and non-nuclear nations due to the expected low cost and higher proliferation
resistance. The high outlet temperatures further allows the plant to be coupled with thermalenergy storage reserves in areas where there is grid instability due to fluctuating renewableelectricity generation. This will allow a more stable energy market and an expansion ofrenewable power.
It is envisaged that one Wasteburning SSR-W will be required to burn the spent fuel andtransmute the long lived actinides for ten thermal spectrum reactors. Extremely high levels oflanthanide contamination are permitted in the fuel of SSR-W. This enables a simple andeconomical pyroprocessing method to be used to convert virtually any spent fuel to fuel.

353

3. Development Milestones
2014
2015

-UK patent granted for use of unpumped molten salt fuel in any reactor.
-Independent capital cost estimate complete-Conceptual design complete and key claims validated by the UK’s National Nuclear Laboratories.

4. General Design Description
Design Philosophy
The entire design philosophy is to reduce plant costs by simplifying the design andeliminating instead of containing hazards. This is done by combining the safety andoperational benefits of molten salts along with conventional reactor components. Risks to the
public are practically eliminated by design, and not merely contained.
The key features of the design are to achieve:• Virtual elimination of the volatile radiotoxic source term under any conceivable
accident, terrorist act or act of war. This would massively reduce the uninsurable
liability (~$100 billion) currently born by national governments in the event of aserious release of radioactivity.
• Deployment of the SSR on smaller sites with smaller emergency planning zones as aresult of this huge reduction in potential for offsite radioactive releases is expected tobe approved by regulators. The probability that this is true is endorsed by the UK’s
National Nuclear Laboratory (NNL) [ 1].
• Economic competitiveness to have a capital cost similar to coal fired power stations
but with substantially lower fuel cost. This is partly as a result of the intrinsic safetycharacteristics and partly due to the absence of pressurised reactor components – seedetails under Plant Economics.
• Modular design with the reactor assembled from road transportable, factory produced
modules creating a single reactor unit of 300MWe. Larger plants of any size canreadily be produced for larger demands which drop the Levelised Cost Of Electricity(LCOE) cost further.
• Fuel assembly form is compatible with IAEA Safeguards procedures used in reactors
today (unlike other molten salt fuelled reactors).
• The output temperature of the reactor is high at 650-700°C. Thermal stresses are
sufficiently low that the proposed standard stainless steels will not suffer creep
degradation at these temperatures. The high output temperature has major advantages
over the low output temperature of PWR reactors:• Use of standard low cost high efficiency superheated steam turbines now
produced on a large scale for combined cycle gas turbine plants instead oflarger, much less efficient saturated steam turbines which are now rarely
produced except for nuclear reactors and which are commensurately more
expensive• Heat storage using proven technology from solar thermal power plants ispossible so electrical generation can be boosted during peak hours while thereactor runs at constant power. This allows carbon free load following toincrease the capacity of renewables that the grid can accommodate.
• SSR process heat has a wider range of commercial uses compared to lowtemperature nuclear steam.
• To design a reactor system that addresses both the long term sustainability of nuclear
energy with the use of thorium and the problem of its resulting long lived radioactive
waste.

354

Power Conversion Unit
A molten salt to steam boiler is proposed to generate steam. The turbine currently being
specified for a 300MWe plant is the Siemens SST5-5000. The air cooled generator is Siemens
SGen5-1200A-2P 118-55. Steam output temperature is 600˚C.
There will either be a larger turbine of 450MWe installed, to be coupled with the solar saltstorage or two turbines of 300MWe and 150MWe. This will depend on the local electricity
grid needs and economics.
Reactor Core
The core is made up of fuel assemblies which sit into a diagrid structure located at thebottom. At the top the assemblies sit in a rail system which forms part of the reactor ‘lid’. The
core is rectangular and the assemblies travel in rows laterally across the core. Each thermal
spectrum assembly contains 37 vented tubes containing the fuel salt surrounded by a solid
block of clad graphite as shown in Figure 2. The coolant passes around the tubes through thecentre of the graphite.

Fuel tube(red)
Coolant channel(grey)
Steel clad graphite(black)Figure 2: Section through fuel assembly showing the graphite, fuel tubes and coolant flow
channel (Reproduced courtesy of Moltex)
The tubes are stainless steel, similar in materials and structures to today’s AGR fuel. The
tubes are 3m long with 2.6m of fuel and 400mm of gas space at the top which has a venting
mechanism as shown in Figure 3 to allow some gaseous fission products to be released into
the coolant salt (most radiologically important fission products are captured in a non-gaseousform). In addition to preventing build-up of pressures within the tubes, the venting
mechanism walls are cooler than the gas itself which causes it to condense on the walls andflow back into the fuel. Fission product gases have time to decay to more stable isotopes
before entering the pool of coolant salt. Most released fission product gases are furthermore
absorbed by the coolant into stable salt forms. Only the noble gases and cadmium migrate
through the coolant salt to the argon containment zone above.
Fuel Characteristics
The thermal spectrum fuel is a molten fluoride salt with a eutectic composition of 33%NaF /30%RbF / 37%UF4 and a melting point of 535˚C. The fuel is prepared in a simple single step
reaction from uranium hexafluoride, as is already produced in many parts of the world. Fuelpellets are melt cast from this mixture requiring none of the complex sizing, sintering and355

machining required for oxide fuel. SSR fuel therefore has the potential to be substantiallycheaper than today’s fuel. Once the assembly is lowered into the tank of molten coolant salt,
the fuel pellets within the tubes will melt. Only then, will they be taken into the core where
they will become critical.

Figures 3: Venting mechanism at the top of each fuel tube causes gaseous fission products tocondense on the walls ensuring a hold up time before it is released to the coolant salt beyond
the tube (Reproduced courtesy of Moltex)
Fuel Handling System
The rectangular core enables lateral fuel shuffling through a rail system without lifting theassembly vertically out of the coolant during a fuel change. The vessel does not need to be
flooded as it does in a PWR. The assemblies in alternate rows are moved in opposite
directions during reload. This result in a pattern where if fresh fuel is inserted in one end, thetwo assemblies either side of it are depleted assemblies ready to be removed - they were
initially loaded as fresh assemblies from the other side of the core. A new fuel assembly isadded to a row every 6 months. At the same time, the assembly at the other end is removed.
New assemblies are added to different rows every few days. This is similar in strategy to
CANDU fuel shuffling and results in a sufficiently flat power profile across the core.
Once an assembly leaves the core it sits at the edge of the tank until it has cooled sufficiently
to be lifted into the argon space above the core to cool down. The fuel quickly solidifies dueto the high temperature gradient between itself and the argon. It can then be removed fromthe reactor and placed into dry cask storage until ready for reprocessing or geologicaldisposal.
Reactivity Control
No excess reactivity needs to be added to the core to compensate for fuel burn up because the
combination of 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, which isreadily compensated by changes in the coolant flow rate.
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 thecore reactivity. The excess reactivity always required at start up is provided either by step bystep addition of the final fuel assemblies with the coolant at normal operating temperature orby adding all the fuel assemblies with the coolant at a higher temperature and then carefully
lowering the temperature. A combination of the two methods is also practical.
356

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Moltex EnergyCountry of origin

United Kingdom

Reactor type

Static Fuelled Molten Salt Reactor

Electrical capacity (MW(e))

300MWe (formed by eight 37.5MWe modules)

Thermal capacity (MW(th))

750MWthExpected capacity factor (%)

91%

Design life (years)

60 years

2

Plant footprint (m )22500

Coolant/moderator

Graphite as an integral part of fuel assemblyPrimary circulation

Forced circulation

System pressure (MPa)AtmosphericoCore inlet/exit temperatures ( C)550˚C - 700˚C

Main reactivity control mechanism

Negative temperature coefficient (net of fuel plus coolant is -5.4pcm/K)

Tank height (m)4m (H)

Tank Size (m)18m (L) x 5m (W)Module weight (metric ton)

<10t per 37.5MWe module excluding salt

Configuration of reactor coolant system

Pool type

Power conversion process

Steam generator to turbine

Cogeneration / Process Heat Capabilities

Yes, possible.

Passive Safety Features:

Large negative temperature coefficients, decay heat removal by passive air
cooling ducts, volatile source term reduced by 106 in accident scenario (dueto the lower partial vapour pressures of volatile fission products in a severe
accident with molten salt fuel compared to a uranium oxide fuel pellet).

Active Safety Features:

Shutdown blade SCRAM + fluoroborate poison

Fuel type/assembly array
Fuel assembly active length (m)Molten salt fuel within vented fuel tubes in a conventional style fuelassembly within graphite.
2.6m

Number of fuel assemblies

1,040 (per 300MWe)

Fuel enrichment (%)<15% LEU

Fuel burnup (GWd/ton)>150GWd/tHM

Fuel cycle (months)~84 months

Approach to engineered safety systems
Number of safety trains

Reactor has multiple inherent safety features. Inherent features will be givena low probability of failure in the Probabilistic Safety Assessment.
Nominally 4 way segregation where applicable

Emergency Safety Systems

Passive Air Cooling Ducts around the tank walls; multiple diverse SCRAMResidual Heat Removal System

Thermal inertia of tank by natural circulation and passive air cooling ducts.

Refuelling outage (days)

On line regular refuelling – maintenance outage under review.

Distinguishing features

Molten salt fuel constrained in conventional, vented fuel tubes. Thermalenergy storage to increase peak electrical capacity.

Modules per plant

8 per 300MWe (no upper limit to number of tanks)

Target construction duration (months)

31 months from start of nuclear concrete

Seismic design

Yes, suspended tank on seismic fixings.

Core damage frequency (per reactor-year)

Expected <10-6 (and the consequence / hazard to the public reduced by 106)

Design Status

Concept, Preparing licensing pack357

Reactor shutdown is provided by shutdown blades with electromagnetic securing of thecontrol blades. The backup method is a coolant neutron poison system based on sodium
fluoroborate which can be released based on a thermally triggered system.
Reactor Containment System
There are no pressurised systems or components within the reactor building. The tank is astainless steel vessel suspended using anti-seismic suspension fixings. The tank is cooled byair cooling ducts around the perimeter. A thin walled stainless steel liner surrounds the walland roof of the argon containment zone. This is surrounded by a ~1m blast resistant concrete
wall which also acts as a biological shield.
A large airlock allows core modules, fresh and spent fuel assemblies to be moved in and outof the core. Argon venting is done irregularly through the airlock.
Reactor Coolant System
The reactor is a pool type reactor so the primary coolant fills the tank. Use of an RbF based
fuel salt means that the coolant must operate between 570°C and 700°C in order to prevent
fuel salt freezing. The coolant salt is therefore 39% ZrF4 / 1% ZrF2 / 60% NaF with a melting
point 500°C.
A secondary coolant salt with additional potassium fluoride to lower the melting point is used
to take the heat from the primary heat exchanger to the boiler tubes. Its composition is 10%
NaF / 48% KF / 42% ZrF4 and it has a melting point of 385°C. The solar salt thermal tanks
are heated through a direct loop into the solar salt tanks. Initial heating of the water/steam
from the condenser is done with a small heat exchanger of solar salt to bring the steam
temperature to above 435°C.
Steam Generator
The boiler tube design is under review but is being designed using knowledge gained fromthe concentrated solar industry. Due to the high melting point of the secondary coolant(385°C), a solar salt of lower melting point is used to do the initial heating of the water/steam
from the condenser before the steam can enter the main steam generator. This is to ensure
there is no freezing of the secondary coolant fluoride salt on the boiler tube walls.
5. Safety Features
Engineered Safety System Approach and Configuration
The molten salts used in the SSR are chemically stable with minimal reactions with air orwater. Specific design solutions and material choices prevents corrosion of the fuel clad (withfuel salt) and plant components with the coolant salt. Use of molten salt fuel with the correct
chemistry further eliminates the hazardous volatile iodine and caesium source terms which
prevents airborne radioactive plumes in severe accident scenarios.
Pool type reactor designs are well established as particularly safe options since naturalconvection of heat from the core is sufficient to remove decay heat. This design element has
been incorporated into the SSR. However, the coolant is chemically non-reactive, under nopressure and has an extremely high boiling point making the pool design even safer thanthose used in other reactors. This single factor separates the SSR from all other molten salt
reactor designs where relatively complex engineered systems are required to assure continued
cooling of the fuel.
The non-volatile radioactive fission products are physically contained within the fuel tubes.
358

In the unlikely event of tube rupture the coolant salt has been selected so as to be misciblewith the fuel salt. If multiple tubes are ruptured, the large tank of coolant provides a meansof massively diluting the fuel rendering the reactor subcritical while maintaining its
containment within the reactor tank without relying on drain tank valves.
Below ground construction - The reactor tank is located below ground level in a steel and concrete

lined pit. This avoids any potential for a physical loss of coolant sufficient to expose the coreand provides protection (in conjunction with above ground concrete shielding) against
aircraft impact or other external factors.
Decay heat removal system
Natural convection of the primary coolant salt will continue in the event of a reactor
shutdown or pump failure. The primary coolant heats up to a point where radiative heattransfer from the tank walls becomes the dominating factor. A finned air duct to atmosphere
exists around the tank walls which can take decay heat away in this accident scenario.
Emergency core cooling system
The core, heat exchangers and pumps are designed so that natural circulation can continue
indefinitely. This ensures the core is continually cooled.
Containment system
See ‘Reactor Containment System’ under Section 4 above. There are no internal pressures.
The primary containment is the tube wall, secondary is the coolant salt (which absorbs fission
products). The third is the tank itself and the fourth is the concrete structure. Above the tankis an argon space which has a stainless steel liner surrounded by a ~1m concrete wall which
serves as the biological shield. The concrete above and below grade is designed to be blast
resistant. The argon containment zone is at a lower pressure than the air within the reactor
building to minimise gas escapes in the event of a breach. However, the composition of thegas within the containment is maintained such that a major release due to massive
containment loss would not release hazardous quantities of radioisotopes. The reactor
building walls are ~300mm thick reinforced concrete. This serves as building structure and as
an initial energy reducer from an external impact of a missile. The building will be aircraft
resistant although the consequence of an aircraft impact is substantially lower than in a PWR
due to the reduced volatile source term.
6. Plant safety and Operational Performances
The design philosophy is such that no operator access is ever required in the main reactor
zone. The substantial reduction in quantity of engineered safety and component systems will
substantially reduce the number of operating staff required. The ramp rates of the plant will
be driven by the steam side, not the nuclear side. Shutdown and maintenance periods will
also be driven by the requirements of the steam turbine.
7. Instrumentation and Control systems
Primary reactivity control will be by the reactivity coefficient of the coolant and fuel. Heatremoval driven by the turbine will control the reactivity sufficiently. There will be neutron
and temperature sensors above the core area and within the coolant. All components aredesigned with the facility to be inspected remotely by visual or mechanical means.
There will be radiation and chemical monitoring of the gas space to ensure no unexpectedfission products are being released to atmosphere (in the event of a leak or opening of the airlock).

359

8. Plant Arrangement

Figure 4: Illustrative site layout for a 300MWe plant (Reproduced courtesy of Moltex)
9. Design and Licensing Status
The focus to date has been on the novel aspects of the reactor including neutronics andchemistry. The majority of the design work has been carried out by various UK consultancies
and universities. The current phase of design is at component level where supplierengagement is taking place to specify each component. The Preliminary Safety Case is inpreparation to initiate licensing in 2017.
10. Plant Economics
Partly as a result of the intrinsic safety and partly due to absence of pressurised reactor
components the overnight cost of the complete nuclear island, including civil works, of a
1GW fast version of the SSR has been estimated independently of Moltex Energy by Atkins
Ltd as being £714 per kW ($1,071/kW) capacity. This costing has been used as the basis for
estimating full plant cost for a 1GWe thermal spectrum SSR-U of £1,892 per kWe($2,838/kW). These are based on UK on-site construction costs, potential cost savings formodularised construction is not included. Substantial prelims and contingencies are included.
This provides a highly credible case for reactors having comparable capital cost to coal fired
power stations but with radically lower fuel cost. Levelised electricity costs are not given at
this stage as they are so heavily dependent on the specific financing mechanisms.
The far simpler design of the Stable Salt Reactor compared to PWRs and AGRs and small
number of safety critical systems should lead to significantly lower O&M costs. The process
for converting enriched uranium hexafluoride into fuel ready for incorporation into fuelassemblies is very much simpler for Stable Salt Reactor fuel than for oxide fuel and is also
therefore expected to be significantly cheaper. Subsidies will not be required as the electricity
produced by the SSR will be competitive with fossil fuels even without carbon tax.
[1] Review of Key Claims of the Stable Salt Reactor, NNL plus other Subject Matter Experts(report available under confidentiality), 2016.
Further information can be found on line at www.moltexenergy.com

360

SmAHTR (Oak Ridge National Laboratory,
United States of America)
1. Introduction
SmAHTR is a deliberately small (125 MWt) fluoride salt–cooled high-temperature reactor(FHR) design concept intended to match the energy requirements of coupled industrial
processes. SmAHTR’s overall purpose is to enable an integrated assessment of the potential
performance of deliberately small FHR designs and to provide guidance to the overall FHR
research and development effort. SmAHTR’s development effort has been in support of the
US Department of Energy (DOE) Office of Nuclear Energy’s Advanced Reactor
Technologies Program. SmAHTR is an early phase concept and therefore is not commercially
viable at this time.

Figure 1: SmAHTR in and near vessel structures (Reproduced courtesy of ORNL)
FHRs are an emerging reactor class that combines attractive attributes from previously
developed reactor classes and power plants. FHRs by definition feature low-pressure liquid
fluoride salt cooling; high-temperature-tolerant, salt-compatible fuel; a high-temperature
power cycle; and fully passive decay heat rejection. FHRs have the potential to economicallyand reliably produce large quantities of electricity and high-temperature process heat while
maintaining full passive safety. Leveraging the inherent reactor class characteristics avoids
the need for expensive, redundant safety structures and systems and is central to making the
economic case for FHRs. Additionally, as a high-temperature reactor class, FHRs canefficiently generate electricity and provide the energy for high-temperature industrial
processes (notably including the production of hydrocarbon fuel). Moreover, hightemperature operation increases FHR compatibility with dry cooling. Figure 1 shows thelayout of the SmAHTR in-vessel structures.
2. Target Application
The particular variant of SmAHTR described in this report is designed for efficient hydrogen361

production using the carbonate thermochemical cycle (CTC) [1] The SmAHTR-CTC design
variant is an evolution of the SmAHTR design concept first described by Greene et al. in2010 [2]. Additional detail about the design is available in ORNL/TM-2014/88.
3. Development Milestones
2010

Initial SmAHTR conceptual design developed

2014

Perform refined SmAHTR concept study focused on high-temperature heat production

4. General Design Description
Design Philosophy
The “new build” strategy being pursued for SmAHTR-CTC has three guiding principles: (1)leverage plant characteristics for maximum economic performance, (2) maximize reliability,and (3) minimize licensing risk. The design intent is to leverage the inherent reactor
characteristics to drive down cost while to the extent possible staying within the structure of
the current US Nuclear Regulatory Commission (NRC) licensing framework. A key elementof the strategy is to rely upon SmAHTR’s strong, inherent safety characteristics to avoid theneed for expensive, redundant safety structures and systems.
Power Conversion Unit
SmAHTR’s hydrogen production plant will implement the CTC. The peak cycle temperature
is not greater than 650°C, and the peak pressure is at most a few atmospheres. In the CTC
process, uranium valence changes drive the decomposition of water to hydrogen and oxygen.
The key innovative step is reacting triuranium octoxide (U3O8) with sodium carbonate(Na2CO3) and steam to generate hydrogen and sodium diuranate (Na2U2O7) at ≤650°C [3].
The reaction is suitable for implementation in a screw calciner. The relatively narrow andvariable energy gap, between uranium’s 6d and 5f atomic orbitals, enables uranium to assume
more than one valence state under relatively mild temperatures and pressures. Shiftinguranium’s oxidation state liberates the hydrogen from the steam. The cycle process steps
necessary to regenerate U3O8 and Na2CO3 from Na2U2O7 are already industrially
implemented in the uranium processing industry. However, the CTC to date has only been
demonstrated at laboratory scale.
Reactor Core & Fuel Characteristics

Figure 2: SmAHTR fuel assembly upper end and partial core cross section (Reproduced
courtesy of ORNL)

362

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

ORNL

Country of origin

USA

Reactor type

Molten Salt Reactor—FHR

Electrical capacity (MW[e])NA

Thermal capacity (MW[th])125

Expected capacity factor (%)>90%

Design life (years)

60 years

Coolant/moderator

FLiBe / carbon

Primary circulation

Forced circulation

System pressure (MPa)AtmosphericCore inlet/exit temperatures (oC)

670/700°C

Main reactivity control mechanism

Negative temperature coefficient; control blade insertion

RPV height (m)9RPV diameter (m)3.5

RPV or module weight (metric ton)

22.5 (empty, no lid)

Configuration of reactor coolant system

IntegralPower conversion process

Carbonate thermochemical cycle

Passive safety features:

Large thermal margin, multiple fission product barriers, multiple barriers
to uncovering fuel, strong negative temperature reactivity feedback,thermally triggered shutdown mechanisms, slowly evolving accidents,
passive natural circulation-based decay heat rejection, low-consequence
fuel handling accidents, structurally robust below-grade nuclear facilities,
non-water soluble fuel, reactor isolation from non-nuclear facilities

Active safety features:

Not necessary

Fuel type/assembly array

Plate TRISO

Fuel assembly active length (m)4Number of fuel assemblies

19

Fuel enrichment (%)

8

Fuel burnup (GWd/ton)

69

Fuel cycle (months)

6

Residual heat removal system

Direct reactor auxiliary cooling system (DRACS)
2 out of 3

Refuelling outage (hours)

8

Design status

Pre-conceptual design

SmAHTR uses tri-structural isotropic (TRISO) coated-particle fuel embedded near thesurfaces of carbon plates. SmAHTR uses the uranium oxycarbide TRISO fuel particles
currently being tested under DOE’s advanced gas reactor fuel development program. The fuelparticles are located near the surfaces of the plates to lower the peak particle temperature by
improving the thermal coupling to the coolant. The plates are configured into hexagonal
assemblies with 18 plates per assembly. SmAHTR’s fuel assemblies will closely resemble
shortened versions of those for the larger advanced high-temperature reactor (AHTR) [4].
The fuel assemblies will be mounted together into a cartridge core using continuous-fibercomposite core support plates. SmAHTR’s cartridge core enables filling of the inter-assemblyvolume with nuclear-grade graphite, resulting in improved neutron utilization, as shown in
Figure 2. The entire core will be lifted out as a single unit for refueling. Each fuel assembly363

includes a molybdenum hafnium carbide control blade. While plate-style TRISO fuel bodies
have not been manufactured previously, the manufacturing process steps of any geometricshape fuel body are nearly identical.
Fuel Handling System
SmAHTR-CTC’s transparent, low-pressure coolant enables refueling to be performed rapidly
using conventional mechanical manipulation technology that is visually guided from above.
To refuel, the reactor is first brought subcritical by inserting the control blades. Once theblades have been inserted, the reactor vessel lid can be removed. The core can then be moved
via an overhead crane system into the used fuel pool. The thermally robust nature of coatedparticle fuel significantly eases the fuel transfer safety design requirements. Even just a few
hours after shutdown, the core can withstand more than 15 minutes of removal from liquid
cooling, during the fuel transfer process, without exceeding its allowable maximum
temperature. SmAHTR-CTC uses a shielded fuel transfer ramp mechanism conceptuallyderived from the Phénix SFR fuel-handling system. Once the reactor vessel top has been
removed, a shielded ramp is positioned (using a rail system) with its entrance over the top of
the reactor vessel. The core is raised out of the vessel and up the ramp using an overheadcrane; and upon reaching the top of the ramp, it is allowed to swing to its opposite side. The
core is then lowered into the used fuel pool, sliding down the far side of the ramp. The corelifting mechanism includes a thermal fuse, so that if at any point during the fuel transfer the
maximum allowable temperature were exceeded, the core would be passively released toslide downward into a cooled configuration. SmAHTR-CTC employs a twin cartridge-core
configuration. A cartridge configuration enables the core to be moved as a single unit,
minimizing the number of mechanical manipulations required for refueling. In a twin coresystem, one core is in the reactor vessel and the other is in the used fuel pool. The twin coreconfiguration enables assembly-level fuel shuffling to be performed while the alternate core
is in operation. SmAHTR’s core needs to be moved as a single unit because of the tightclearances of its structural elements. Some radiation-induced mechanical distortion will occur
during operation. The mechanical distortion will prevent the removal of individual fuelassemblies from the top. The core internal graphite pieces will be replaced as the fast neutron
fluence causes excessive mechanical distortion (every few years). Separating the fuel fromthe graphite moderator enables separate disposal of the in-core graphite and the fuelassemblies, decreasing the volume of high-level waste. SmAHTR-CTC maximizes its
neutronic efficiency by maximizing the amount of carbon in the core while minimizing theamount of FLiBe.
Reactivity Control
SmAHTR has two primary shutdown mechanisms: (1) control blades and (2) poison salt
injectors. SmAHTR does not have any threshold phenomena that would occur if the primarycoolant temperature were to exceed the design point by several tens of degrees for hundredsof hours. Consequently, the temperature rise can be employed to passively initiate shutdown.
If SmAHTR’s core outlet temperature were to exceed its design limits, the control blades
would be passively inserted as a result of melt point fuses in the control blade linkages. The
control blades are jointed to provide sufficient flexibility to enable insertion if their guidetubes become distorted. The control blades are also actively engaged to the drive motors.
Upon loss of power, the control blades would disengage from the drive mechanisms and drop
into the core. The poison salt injectors have a single-ended cylinder and plunger format.
Pressurized inert gas is maintained on the closed end of the cylinder. A sliding disk divides
the cylinder volume into open and closed ends. The open end of the cylinder is filled withpoison salt powder (EuF3 or GdF3). A melt point alloy holds a lid onto the open end of the364

cylinder during normal operations. The poison salt accumulator is located in SmAHTR’slower plenum. If the lower plenum temperature exceeds the set point value, the melt point
trigger releases and the poison salt is injected into the core inlet. The melt point releasemechanism can also be triggered by the plant operators via embedded electrical heaters. Boththe primary and secondary shutdown mechanisms provide substantially more negative
reactivity worth than would be necessary to shut down the reactor. The reactivity worth of thecontrol blades is sufficient to shut down the reactor with the most reactive blade stuck out of
the core at all points during the fuel cycle. As the intended refueling interval is only 6months, blade insertion capability will be confirmed frequently. The solubility of the reserve
shutdown poison salts is much greater than the potential concentration in the lower vessel
plenum, providing confidence that the salts would dissolve completely in the primary coolant
and rapidly bring the reactor subcritical.
Reactor Pressure Vessel and Internals
SmAHTR employs an Alloy 800H reactor vessel with an Alloy N lining. The fueled core andthe radial reflector are contained within a 2-cm-thick core barrel made of carbon-carbon
composite that separates the up-flow from the downcomer. The inside of the core barrel
facing the core has a thin plating (1 cm) of boron carbide to reduce the neutron flux on thereactor vessel. The downcomer is sized so that its transverse flow area is at least twice the incore flow area.
5. Safety Features
The safety characteristics of FHRs arise from fundamental physics and well-designed,
constructed, and maintained structures, systems, and components (SSCs). Until an approved
set of reactor-class-specific general design criteria, as well as design basis accident (DBA)
sequences, is available, a combination of engineering judgment and the safety characteristics
of FHRs will be relied upon to guide SmAHTR-CTC design. FHRs feature full passive safetyand do not require any active system or operator response to avoid large off-site release forany DBA or evaluated beyond-DBA, including earthquake, tsunami, large commercial plane
impact, or permanent station blackout. More specific information about SmAHTR-CTC’ssafety characteristics is provided in the following paragraphs.
Large Thermal Margin To Fuel Failure
The silicon carbide barrier layer of properly operating coated-particle fuel will remain intact
several hundred degrees above SmAHTR-CTC’s peak fuel temperature, retaining nearly allof the fission fragments. The primary FLiBe coolant will not boil until it reaches
temperatures over 1400°C. SmAHTR’s core outlet temperature is 700°C. Additionally, theviscosities of fluoride salt coolants decrease with increasing temperatures, resulting inincreased coolant flow to the hottest fuel, thereby reducing fuel hot spots. Further, while thepower density of FHR cores is above that of high-temperature gas-cooled reactors (HTGRs),
FHRs contain larger numbers of fuel particles, resulting in low average particle power (26.5mW/particle for the SmAHTR-CTC compared with ~150 mW/particle for the nextgeneration nuclear plant HTGR design). This limits the radiochemical challenge to the siliconcarbide containment layer within the fuel.
Multiple Fission Product Barriers
As described in the previous section, SmAHTR-CTC features four independent containment
layers along with a substantial radionuclide source term reduction due to the core’s
immersion in fluoride salt. Additionally, the low system pressure prevents the development ofa significant driving force, which would cause the dispersion of radionuclides in the event ofbarrier failure.
365

Multiple Barriers To Prevent Uncovering FuelThe reactor vessel does not include any penetrations below ~20 cm above the top of the direct
reactor auxiliary cooling system (DRACS) heat exchangers. Fluoride salts have highvolumetric heat capacity, and a large volume of salt is maintained over the core, providing asubstantial thermal margin. Further, owing to the high melting point of the primary salt
coolant, small leaks in the reactor vessel are expected to self-seal as the coolant freezes uponleaking. Although the fuel is slightly buoyant in the coolant, multiple upper core support
structures, including the upper core support plate, the weight of the control blade, and thecontrol blade guide tube, all maintain the fuel below the surface.
Strong Negative Temperature Reactivity Feedback
The strong overall reactor negative reactivity temperature feedback will prevent reactivity
insertion accidents from causing fuel damage. The negative Doppler fuel temperature
coefficient for fresh fuel is similar in magnitude to that of a typical pressurized water reactor.
The coolant temperature feedback coefficient is also negative, but it is one order ofmagnitude smaller than the negative Doppler feedback, while the moderator temperature
coefficient is near zero. The >700 K margin to coolant boiling, combined with the deep
coolant pool, makes significant voiding of the core highly improbable. The combination ofvoid and temperature coefficients is always negative under these circumstances, thus
avoiding the potential for fuel damage reactivity accidents. Bubbles within the primarycoolant flow (caused by, for example, gross failure of the reserve shutdown system piping)would be a DBA that would cause local fuel heating as the bubbles rise through the core.
However, the large thermal margin to fuel failure, the rapid Doppler feedback, the limited
duration of bubble passage, and the longer neutron path length in graphite-moderated cores
alleviate the potential for bubble entrainment to damage the fuel. Additional safety
evaluations are necessary to better assess the heat transfer consequences of the events under
this scenario.
Thermally Triggered Shutdown Mechanisms
SmAHTR-CTC’s most thermally sensitive component is its reactor vessel. Exceeding thereactor vessel’s design temperature for long periods of time would increase its creep.
However, creep is not a threshold process, and SmAHTR-CTC’s reactor vessel creep budget
would allow for hundreds of hours of operational time at several tens of degrees above
normal. The ability of the vessel, core, and coolant to tolerate temperature rises enables theuse of thermal triggers to provide negative reactivity feedback. SmAHTR-CTC’s control
blades feature thermal fuses (melt point alloys) located in the upper plenum that are designed
to release the blades in the event of an over-temperature accident without SCRAM. Also,
SmAHTR’s secondary shutdown poison salt accumulators will be set to inject poison salt into
the core inlet plenum in the event of more-severe temperature excursions.
Slowly Evolving Accidents
SmAHTR-CTC has no credible primary coolant-phase-change accidents. SmAHTR-CTC
also has a much lower power density core than modern light water reactors (LWRs) (~5.3
MW/m3 vs ~110 MW/m3), resulting in significantly slower transient response. The power
density is calculated to be 9.4 MW/m3 if the carbon between the fuel assemblies is notconsidered to contribute to the core’s volume (referred to as “power density in fuelledvolume”). Further, the large volume of salt (the volumetric heat capacity of FLiBe is 4670
kJ/m3-K at 700°C) and graphite in the reactor vessel provides for substantial local thermal
storage to provide time for recovery operations.
Passive, Natural Circulation-Based Decay Heat Rejection

366

SmAHTR employs natural circulation DRACS-type decay heat rejection to the local air.
Fluoride salts have large coefficients of thermal expansion that, combined with their large
volumetric heat capacities, enable excellent natural circulation cooling. Similar naturalcirculation-driven emergency heat rejection is also provided to the used fuel storage pool.
The DRACS are modular and independent, enabling the system to continue to function
following the failure of individual units. SmAHTR-CTC’s design calls for three independent
DRACS and is intended to withstand DBAs with only two out of three systems in operation.
Sufficient decay heat rejection is provided to avoid vessel failure in the event of complete,permanent station blackout accompanied by the loss of a single DRACS. In addition,
SmAHTR-CTC will not require a dual interconnection to the grid, as the plant does notdepend on the availability of electrical power to reject decay heat. SmAHTR-CTC does not
require an emergency core cooling system or high-volume, safety-grade coolant makeup
systems. Significantly lowering the primary coolant level in the vessel requires both primaryand guard vessel failures, as no other mechanisms have been identified to remove substantial
amounts of coolant from the vessel.
Low-Consequence Fuel-Handling Accidents
SmAHTR-CTC’s core is slightly buoyant. However, each fuel assembly includes amolybdenum alloy control blade to provide sufficient ballast to sink the fuel in the event thatthe core is dropped during movement. Liquid fluoride salts are transparent, enabling fuelmovement to be visually guided, thus substantially reducing the risk of unanticipatedmechanical impacts while enabling visual confirmation of mechanical motion. Furthermore,since only the weighted core sinks slowly in the coolant, mechanical damage to the core or
any other reactor structures as a result of its being dropped would not be anticipated.
Structurally Robust, Below-Grade Nuclear Facilities
FHR buildings and used fuel storage pools are intended to be located below grade, providing
shielding from aircraft impact and wind/waterborne missiles. The DRACS cooling towers are
spatially separated to avoid simultaneous failure of both as a result of a single aircraft impact
and are structurally robust to minimize their vulnerability to likely natural events (e.g., large
tornados or earthquakes).
Non-Water-Soluble Used FuelCarbon-enveloped coated-particle fuel is not water soluble, enabling used fuel assembly
emplacement in shallow (<1 km) dry wells for eventual (after 6–12 months in an incontainment used fuel pool), local, intermediate-term, retrievable storage with a very low
probability for radionuclide escape into the surrounding environment. The first layer ofcontainment remains with the fuel after it is unloaded from the core, and the silicon carbide
containment layer is not anticipated to fail in the more benign used fuel storage environment.
The envisioned local used fuel storage dry wells (boreholes) closely resemble cased andcemented petroleum wells, which can effectively reject decay heat to the surrounding rock as
a result of their large surface-to-volume ratio.
Reactor Isolation From Non-Nuclear Facilities
SmAHTR-CTC’s low-pressure liquid-salt intermediate loop prevents accidents in the CTCfrom adversely impacting the nuclear plant. The relatively low cost of the intermediate loop
per unit of length enables separation of the reactor building from the hydrogen production
facility by substantial distances (>100 m). The reactor will also be separated from potential
pressure waves propagating down the intermediate loop piping by rupture disks, or otherconventional pressure relief mechanisms, so that it is not credible for a disturbance (includinghydrogen explosions) within the CTC to propagate to the nuclear island.

367

6. Decay Heat Removal
SmAHTR-CTC’s normal core decay heat removal is via the intermediate cooling loop to thebalance-of-plant heat rejection system. During maintenance shutdowns, SmAHTR-CTC’score will be removed to the used fuel cooling pool. Decay heat will be removed via the three
independent DRACS cooling loops under loss-of-forced-flow accident conditions. The
DRACS cooling loops transfer heat from the DRACS heat exchangers located within thereactor vessel to the natural draft heat exchangers (NDHXs) located at the bases of chimneys
external to the reactor building. Each NDHX is located within its own chimney to avoid
common vulnerability to a single external impact. During normal operation, each NDHX iscontained within an insulated shell to minimize the parasitic heat loss, and resultant potential
for freeze-up, and to enable tritium capture within the NDHX shell. Each DRACS loop issized to provide ~0.5% of full-power heat rejection under fully developed flow conditions at
700°C. Only two of the three DRACS loops are required to protect against an unacceptabletemperature rise following or during a protected loss-of-forced-flow accident. The ultimate
heat sink is outside air. Used fuel pool normal-condition cooling is provided via a pumped
salt loop to a forced convection heat exchanger coupled to the outside air. Safety-grade heatrejection is provided for the used fuel pool through a two-phase, reduced-pressure, liquidmetal thermosyphon tube wall with the hot ends of the tubes dipped into the used fuel pooland the condenser section above the top of the containment structures.
7. Plant Economics
SmAHTR-CTC’s objective is to economically and safely produce hydrogen and thereby
maximize the return on investment to the plant owners. A key element in ensuring
advantageous economic performance is minimizing the risk premium element of the interest
rate paid during construction. The interest rate risk premium is determined based upon how
confident a lender is that an FHR plant will (1) be constructed at planned cost and schedule,(2) obtain regulatory approval to operate, and (3) operate reliably for many years. SmAHTRCTC’s design is not yet mature enough to develop a “bottom-up” cost model. Consequently,
the economic evaluation is limited to comparison with similar technologies. Nearly all
aspects of FHRs will either be less costly than or have costs equivalent to those of LWRs.
The major exception to the comparative cost advantage is the markedly higher cost of the
primary coolant. Initial FHRs will also incur significant first-of-a-kind expenses. For
example, SmAHTR’s initial fuel load of TRISO coated-particle fuel will cost much morethan an equivalent LWR fuel because of the need to construct new manufacturing facilities.
SmAHTR is a high-temperature reactor, which would enable it either to have higher thermal
efficiency for electricity production or to produce high-temperature process heat. SmAHTRCTC’s design is optimized to provide high-temperature process heat, avoiding initialcompetition with mature technology.
[1]
[2]
[3]
[4]

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J. L. Collins, et.al., Carbonate Thermochemical Cycle for the Production of Hydrogen,
US Patent 7,666,387, B2, February 23, 2010.
S. R. Greene, et.al., Pre-Conceptual Design of a Small Modular Fluoride Salt-CooledHigh Temperature Reactor (SmAHTR), ORNL/TM-2010/199, December 2010.
J. Ferrada, et.al., “Carbonate Thermochemical Cycle for the Production of Hydrogen,”
National Hydrogen Association Annual Conference and Hydrogen Expo, Columbia,
South Carolina, March 30–April 3, 2009.
V. K. Varma, et.al., AHTR Mechanical, Structural, and Neutronic Preconceptual
Design, ORNL/TM-2012/320, Oak Ridge, TN, September 2012

Liquid Fluoride Thorium Reactor (Flibe
Energy, United States of America)
1. Introduction
The liquid-fluoride thorium reactor (LFTR) design by Flibe Energy is a graphite-moderated,thermal-spectrum reactor with solutions of liquid fluoride salts containing both fissile andfertile materials. Thermal power generated from nuclear fission would drive electricalgeneration 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 in which nuclear fission reactions can take place. The ionically-bonded natureof the salts prevents radiation damage to the mixture and allows for operation at hightemperature yet at essentially ambient pressure.

Figure 1: View of the liquid-fluoride thorium reactor (LFTR) plant design(Reproduced courtesy of Flibe Energy)
The high operational temperatures of the fluoride salts (500-700°C) make them excellent
candidates for coupling to a closed-cycle gas turbine power conversion system (PCS). The
supercritical carbon dioxide gas turbine employing the recompression cycle is proposed andcan generate electricity at high efficiencies (approximately 45%).
The LFTR design has a two-region core (feed / breed) and utilizes a closed fuel cycle basedon thorium. The reactor vessel incorporates two plena with a central active core region andthe outer blanket area, both filled with fluoride salt. The thorium-232 in the blanket region is
ultimately converted to uranium-233 through neutron capture and beta decay. The chemical
processing system is used to separate and re-introduce the fertile and fissile material to thetwo fluoride fuel-salt streams respectively. Utilizing thorium fuel in a thermal neutron
spectrum, the reactor is able to extract almost all the energy content thus assuring practicallyunlimited thorium resources and the associated insignificant basic fuel costs.

369

2. Target Application
Develop a power-generating nuclear reactor that will produce electrical energy at low cost by
efficiently consuming thorium.
3. Development Milestones
2015

Completion of EPRI-funded study of LFTR design

4. General Design Description
Design Philosophy
The objective of the liquid-fluoride thorium reactor (LFTR) design proposed by Flibe Energy[1] is to develop a nuclear power plant that will produce electrical energy at low cost. Byutilizing thorium fuel in a thermal neutron spectrum, the reactor is able to utilize the energy
content of the thorium at a very high efficiency, approaching 100%, at which point the Earth's
thorium resources practically becomes unlimited. Some of the main principles followed in thedesign are (i) inherently safe, with a no meltdown and non-pressurized core; (ii) simplicity, tohave an intrinsically stable and self-regulating design; (iii) fuel efficient, and (iv) have thepotential to produce far less waste. The general plant layout is shown in Figure 1.
Nuclear Steam Supply System
The nuclear heat supply and power conversion system is shown in Figure 2. The individual
systems are described in more details below followed by other system design descriptions.
Reactor Core
The reactor vessel functions to hold fuel salt, blanket salt, and moderator material together insuch a way so as to maintain a critical configuration at the temperatures and thermal power
levels required. In addition, it incorporates reactivity control mechanisms both active andpassive. 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 iscirculated through the reactor.
The reactor vessel design incorporates several safety functions. In many accident events, afreeze valve, which form part of the vessel and primary loop system, melts and allows fuelsalt to drain from the primary loop and the reactor vessel into the drain tank. The separationof the fuel salt from the solid graphite moderator retained in the reactor vessel, assures a
subcritical configuration can be established in the drain tank.
The internal graphite structures needs to be replaceable since it subjected to a fast andthermal neutron flux that is greatly in excess of that which will be experienced by the metallic
reactor vessel itself, and the replacement of these graphite structures will enable the reactor
vessel to continue to operate and serve its function.
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 thorium-232 via nuclear beta
decay, first to protactinium-233 and later to uranium-233. Both the protactinium and theuranium are chemically removed from the blanket salt mixture and introduced into the fuelsalt mixture in the reactor to fission. The fission products are later chemically removed fromthe fuel salt and in some cases separated and purified before final disposition.

370

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

Flibe EnergyCountry of origin

United States

Reactor type

Molten Salt Reactor

Electrical capacity (MWe)250

Thermal capacity (MWth)600

Expected capacity factor (%)

90+

Design life (years)

Undetermined

2

Plant footprint (m )To be determined

Coolant/moderator

Graphite moderator, LiF-BeF2-UF4 fuel salt

Primary circulation

Forced circulation

System pressure (MPa)AmbientoCore inlet/exit temperatures ( C)

500 / 650

Main reactivity control mechanism

Negative temperature coefficient; control rod insertion

RPV dimensions and weight

To be determined

Configuration of reactor coolant system

Loop

Power conversion process

Supercritical CO2 gas turbine

Cogeneration / Process Heat Capabilities (High / Low

Example: Yes, possible with different configuration

Passive Safety Features:Freeze plug releasing to drain tank; large negative

Active Safety Features:

Motor-driven control rods for shutdown

Moderator type/assembly array

Graphite prisms, triangular pitch

Moderator assembly active length (m)To be determined

Number of moderator assemblies

To be determined

Fuel enrichment (%)

Not applicable, uses uranium-233 fuel derived from Th

Fuel cycle (months)

Continuous refueling from U-233 produced in blanket

Approach to engineered safety systems

Passive safety implemented throughout

Number of safety trains

Not applicable

Emergency Safety Systems

Not applicable

Residual Heat Removal System

Drain tank in thermal communication with environment

Refuelling outage (days)

Not applicable

Distinguishing features

Complete consumption of thorium resource for energy

Modules per plant

4-6

Target construction duration (months)

To be determined

Seismic design

To be determinedCore damage frequency (per reactor-year)

Not applicable

Design Status

Concept371

Figure 2: Reactor and primary loop, intermediate loop, power conversion system, andexternal cooling system simplified flow diagram (Reproduced courtesy of Flibe Energy)
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 transfer its heat to the coolant salt. The
primary pump provides the necessary forced circulation. The primary loop system begins andends with its connection to the reactor vessel and includes the primary pump, the PHX, thebubble injection system, and the fuel salt drain tank and its associated external coolingsystem.
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 (orintermediate heat exchanger, IHX), the coolant salt drain tanks, and the pressure relief(blowout) valves.
The intermediate loop also isolates the primary loop from the high pressures of the PCS usingpressure relief valves. The isolation is an important safety function. In case of a failure in thehigh pressure PCS it will prevent the transmittal of high pressure back through the coolant
salt to the primary loop. The primary loop is not designed for high pressures and without
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 the372drain of the primary loop fluid contents into the fuel salt drain tank (also see passive
shutdown and heat removal later).
The intermediate loop also serves another practical purpose. Since cooling fuel salt with acoolant salt is more compact and effective than cooling fuel salt directly with a gas, the PHX
is much smaller and operates at low pressures. It also reduces the fuel salt inventory of the
primary loop which reduces the amount of fissile material needed for a given power rating.
Power Conversion System
The function of the PCS is to convert the maximum amount of enthalpy contained in theheated 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 therecompression 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 thehigh-temperature and low-temperature recuperators. It also includes the main turbine, themain compressor, the recompressor, and the electrical generator. The PCS interfaces with theintermediate loop through the gas heater, and interfaces with the external cooling systemthrough the gas cooler.
The PCS and carbon dioxide working fluid in the cycle provides a final barrier to tritium
release into the environment. Tritium generation is an inevitable consequence of using
lithium and beryllium in the salt mixture and thus the PCS also includes a tritium removal
system.
External Cooling System
The function of the external cooling system is to reject the heat that was not converted toshaft power in the PCS to the environment in an acceptable manner. The design shall also
prevent the transmission of tritium to the outside environment.
Fuel Handling System
New fissile material can be easily added or removed without resorting to changing thechemical nature of the solvent, and this allows overall reactivity to be held very close to theminimum amount needed to achieve criticality. The fuel salt control is done by the chemical
processing system.
Chemical Processing System
The main function of the chemical processing system is to remove uranium and protactinium
from the blanket salt and to return uranium to the fuel salt. Its secondary function is toremove fission products from the fuel salt and to further process them into acceptable forms.
The safety-related functions of the chemical processing system mainly involve the safe
handling of highly radioactive materials. Drain tanks and cooling systems must be provided
for each reaction vessel at each stage of processing through the system. Unfortunately highly
chemically reactive gaseous fluorine and hydrogen is needed, to be created just as they areneeded from an electrolytic cell using anhydrous hydrogen fluoride as the feed.
Various fission products disperse into the fluid streams of the chemical processing system andsome must be handled differently than others. A class of fission products including selenium
and tellurium will migrate with gaseous hydrogen and hydrogen fluoride and are handled in a
potassium hydroxide neutralization cleanup system. Other fission products are removed fromthe fuel salt in a reductive extraction column and will exist in a metallic state in bismuth. The
high chemical potential of a metal form means that these fission products will need to be373

oxidized and placed in a disposal form before shipment from the site. The small amounts ofmaterial produced means that these disposal plans will not constitute a major issue withreactor operations.
Substantial development work will be needed to prepare for long-term operation of thechemical processing system.
Offgas Handling System
The function of the offgas handling system is to provide a sufficient holdup volume for xenon
and krypton generated in the fission reaction, allowing all of their radionuclides to decay toother forms with the exception of krypton-85, which has a 10-year half-life.
Xenon and krypton (and to some degree tritium) are the most mobile radioactive terms in thereactor system. Tritium is subject to chemical reactions but xenon and krypton are not.
Fortunately, with the exception of krypton-85, all of their radionuclides are short-lived and aholdup of roughly thirty days is sufficient to allow them to decay to non-mobile forms likecaesium, rubidium, strontium, and barium.
The offgas handling system utilizes the fuel salt decay tank as a primary storage volume,
allowing the initial and most intense stages of decay to take place there. The passive coolingsystem of the fuel salt decay tank is utilized to cool the noble gases, providing a continuous
test of the efficacy of this crucial subsystem. After initial cooling in the decay tank, gaseous
xenon and krypton in a stream of helium pass into a long piping arrangement filled withcharcoal (that adsorbs these gases) and cooled by water that provides sufficient holdup
volume over time.
After all radioisotopes of xenon have decayed away, the remaining gas stream iscryogenically cooled to separate stable xenon from helium and krypton. Xenon is bottled andcould be sold at this stage. Krypton is also bottled and stored because of the continuing slow
decay of krypton-85. Helium is recycled and returned to the gas handling system.
Reactivity Control
The reactor vessel accommodates passive and active control rod systems which also haveimportant safety functions. The blanket salt held within the reactor vessel is a strong neutron
absorber, and a blanket salt leak from the reactor vessel could lead to the reduction in theblanket salt inventory contained in the reactor vessel, increasing reactivity by removing aneutron-absorbing medium. To compensate for this introduction of positive reactivity, aseries 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 theserods, allowing them to slide down into the core and add negative reactivity to replace andovercome the negative reactivity lost from 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 system present 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 thecore due to their strong negative reactivity.
An active set of control rods, of a more conventional design, would also be present in thereactor vessel and would serve a safety function, allowing the operator to control thereactivity level of the reactor. These rods, which would comprise a smaller and less potent
source of negative reactivity, would be clustered near the center of the core and provide finer
control over reactivity levels. Other possible designs are also considered.
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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.
At present, it is anticipated that the reactor vessel will incorporate a small heat exchanger
exclusively meant for cooling the blanket salt, with the blanket salt flowing throughout thecore under the driving force of natural circulation. This heat exchanger has an importantsafety function in that it cools blanket salt which contains short-lived thorium-233, asignificant heating term in the fluid that cannot be chemically removed. If the reactor shut
down or if blanket salt chemical processing was terminated for any reason, the reactor vessel
would also have to accommodate the heating generated by protactinium decay, but thorium233 decay would end relatively quickly in this case, since thorium-233 only has a 22-minutehalf-life.
5. Safety Features
The properties of fluoride salts offer LFTR enhanced safety characteristics. The fluid salt in
the core is not pressurized, thus eliminating the fundamental driving force that can releaselarge amounts of radioactivity to the environment. The notion of a “meltdown” leading toreactor failure becomes irrelevant in a reactor designed around the use of liquid fuels. The
reactivity of the reactor is self-controlling because any increase in the reactor’s operating
temperature results in decrease in density of the fuel salt in the core and a reduction of reactor
power, thus inherently stabilizing the reactor without the need for human intervention or
backup systems.
The reactor is designed with a simple salt plug drain in the bottom of the core vessel tocompletely shut down the core through gravity. If the reactor should lose power or need to be
powered down for any reason, the salt plug is simply allowed to melt and the fluid salt todrain into a passively cooled containment vessel(s) where decay heat is readily removed. This
simple feature prevents accidents or radiation releases due to lack of cooling. It also provides
a convenient means to shut down and restart the reactor quickly and easily.
The safety function of several of the systems was already described above. Below a few
additional safety features of the LFTR are highlighted or further explained.
Fission Product Retention
The integrity of the reactor vessel plays an important role in minimizing radiation hazards byconfining 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 radiologicalconcern, form stable fluoride salts that are retained in the overall mixture under all
conditions. Fission products gases, whose removal is important from a performance andsafety basis, are easily separated from the fluid mixture and allowed to decay to stability in a
separate system.
Passive Shutdown and Heat Removal
An important safety function is embedded in the primary loop and is activated when thereactor overheats or loses its coolant flow. A freeze valve is integrated into the primary loopthat is maintained frozen by an active coolant system. When this coolant is lost or if thetemperature of the system exceeds its cooling capability, the freeze valve fails open and thefuel salt drains out of the primary loop and out of the reactor vessel into the fuel salt drain
tank.
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The fuel salt drain tank is integrated with a separate cooling system that is passively
connected to 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 towater, and do not undergo vigorous chemical reactions with air or water in contrast to manyliquid metals.
The components of fluoride salt mixtures have both desirable and undesirable aspects, andthe two most important are lithium-7 fluoride and beryllium fluoride. The two natural
isotopes of lithium must be separated from one another since lithium-6 (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 andreactive. But the fact that fluoride salt mixtures are processed in a salt form rather than being
dissolved into an aqueous solution mitigates issues of accidental 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 moststructural metal alloys, 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 othermetallic constituents. Fluoride salts moderate neutrons sufficiently on their own to preventthe formation of a truly fast neutron spectrum, but are still insufficiently effective to generatea thermal neutron spectrum. Thus separate moderator materials are necessary for the reactor
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 thegraphite 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 fromdimensional distortion over time, but this effect can be quantified and compensated for inreactor design.
6. Plant Arrangement
The plant arrangement has been shown in Figure 1. 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. Plant Economics
The economic performance of the plant has not yet been modelled in sufficient detail but allindications point towards its performance being strongly competitive with other clean energy
sources.
[1] For more information visit: http://flibe-energy.com/

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Mk1 PB-FHR (UC Berkeley, United States)

1. Introduction
The Mark 1 Pebble-Bed Fluoride-Salt-Cooled High-Temperature-Reactor (Mk1 PB-FHR) isa small, modular graphite-moderated reactor. FHRs are differentiated from other reactor
technologies because they use high temperature, coated particle fuels, and are cooled by thefluoride 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) for base-load electricity
generation. Figure 1 shows the flow schematic for the Mk1 PB-FHR.

Figure 1: Mk1 PB-FHR flow schematic (Reproduced courtesy of UC Berkeley)
2. Target Application
The Mk1 PB-FHR is designed to produce 100 MWe of base-load electricity when operated
with only nuclear heat, and to increase this power output to 242 MWe using gas co-firing forpeak electricity generation. This provides a new value proposition for nuclear power to earn
additional revenues by providing flexible grid support services to handle the ever increasing
demand for dispatchable peak power. This is in addition to traditional base-load electrical
power generation.
3. Development Milestones
To be developed
4. General Design Description
Design Philosophy
The Mk1 PB-FHR is designed with advanced passive safety features and intrinsic fuel andcoolant properties which make the consequences of severe accidents commonly studied forlight water reactors much easier to manage. Passive safety mechanisms include natural
circulation decay heat removal activated by a passive check valve in accident conditions andbuoyant control rods for emergency shutdown without operator intervention.

377

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 heattransfer, and high retention of most fission products. These characteristics are in addition toreasonably low neutron capture probability, and good neutron moderation capability.
The Mk1 design implements several innovative features: The Mk1 PB-FHR does not use anintermediate coolant loop and instead directly heats the power conversion fluid. It eliminates
the conventional reactor guard vessel used in sodium fast reactors and instead uses a
refractory reactor cavity liner system. Furthermore, all components for the Mk1 design arerail-transportable.
Reactor Core
Figure 2 shows a cross section of the Mk1 reactor vessel, internals, and core. 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 outside that reduces the fast-neutron fluence to the outer fixed radial graphitereflector sufficiently for it to last the life of the plant.

Figure 2: The Mk1 PB-FHR reactor vessel (Reproduced courtesy of UC Berkeley)
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 transferfrom the fuel and results in lower average fuel temperature. The center graphite reflector
internals need to be replaced periodically due to radiation damage. Its size is much shorterand smaller than center reflectors proposed for helium-cooled pebble-bed reactors, making it
simpler to design for replacement and for seismic qualification.
Power Conversion Unit
The 236-MWth 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 MWe under base-load operation,
and with natural-gas co-firing to rapidly boost the net power output to 242 MWe to provide
peaking power. The Power Conversion Unit consists of the reactor core, two coiled tube air
378

heaters (CTAHs) to transfer heat from the main salt to pressurized air, a heat recovery steam
generator system (HRSG), steam condenser, and the GT. During normal operation, theprimary coolant relies on forced circulation.
Fuel Characteristics
Figure 3 shows the 3-cm diameter spherical pebble fuel elements. 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 U-235 to 19.9% and encapsulated inside 4,730
coated particles. The very low circulating power for the coolant in salt-cooled reactors,
compared to helium-cooled reactors, makes it practical to use smaller pebbles. This smallpebble 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. Four Mk1 pebbles can provide electricity for a full year toan average U.S. household (approximately 11 MWe-hr in 2011). This is equivalent to 8.1
tons of anthracite coal, or 17 tons of lignite coal, if produced in a coal fired plant.

Figure 3: A PB-FHR pebble fuel element (Reproduced courtesy of UC Berkeley)
Fuel Handling System
The design includes pebble injection channels integrated into the inner surface of the reactor
core barrel. The pebble injection channels follow the curvature of the bottom of the reactor
vessel before discharging pebbles through the lower center reflector support structure. The
Pebble Handling and Storage System (PHSS) will function to ensure the sustained fission
reaction in the core through the addition and removal of fuel and graphite pebbles. Space has
been allocated in the upper core internals and in the reactor building for key PHSS
subsystems and components including the core unloading devices, pebble burnup
measurement system, and pebble canister transfer system, but detailed design has not yet
been completed.
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 thebuoyant rods will insert if the reactor coolant temperature in the control-rod channel exceeds
615°C, the buoyant stability limit. The design also uses shutdown blades that can insert
directly into the pebble bed for reserve shut down. In the event that electrical power isinterrupted to the drive mechanisms for the motors of the control rod and shutdown blade
cable drums, they will insert and shut the reactor down.

379

MAJOR TECHNICAL PARAMETERS
Parameter

Value

Technology developer

University of California, Berkeley

Country of origin

United States of America

Reactor type

Fluoride-Salt-Cooled High Temperature Reactor (FHR)

Electrical capacity (MW(e))

100 MWe

Thermal capacity (MW(th))

236 MWth

Expected capacity factor (%)>90

Design life (years)

60

Plant footprint (MWe/acre)

16.1

Coolant

Li2BeF4 (flibe)

Moderator

Graphite

Primary circulation

Forced circulation

System pressure (MPa)0.3oCore inlet/exit temperatures ( C)

600 / 700

Main reactivity control mechanism

Negative temperature coefficient; control rod insertion

RPV height (m)12

RPV diameter (m)3.5

RPV weight (metric tonnes)

62

Configuration of reactor coolant system

Pool type

Power conversion process

Nuclear air-Brayton combined cycle (NACC)

Cogeneration / Process Heat Capabilities (High / Low T)

Yes, possible with different configuration

Passive Safety Features:

Large negative temperature coefficients, large heat capacity

Active Safety Features:

Yes, control rod SCRAM; Turbine trip

Fuel type/assembly array

TRISO particles in graphite pebble matrix / Pebble bed

Active height of fuel region (m)2.5

Number of fuel pebbles in core and defueling chute470000Fuel enrichment (%)

19.9

Fuel burnup (MWd/kgHM)

180

Fuel core residence time (months)

2.1, average of 8 passes to achieve full burn up

Approach to engineered safety systems

Passive

Emergency Safety Systems

Control rod drive mechanism, shutdown blades

Residual Heat Removal System

Direct reactor auxiliary cooling system (DRACS)

Refuelling outage (days)

0 days (online refuelling)

Distinguishing features
Modules per plant

Large fuel and coolant thermal margin to damage, hightemperature operation, NACC with peaking, passive safety
12

Target construction duration (months)

24

Design Status

Pre-Conceptual Design

Reactor Pressure Vessel and Internals
To enable near-term deployment, the Mk1 design uses a core barrel and other core internal
structures fabricated from the same metallic material as the reactor vessel and main salt
piping. The outer radial reflector blocks are aligned and held against the metallic core barrel
using a system of axial alignment ribs and radial retaining rings quite similar to designs

380

originally developed for the Molten Salt Breeder Reactor (MSBR) project. The use of
metallic core internal structures, rather than advanced ceramic composites, simplifies
fabrication and licensing for the Mk1 design.
Steam Generator
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 design considerations are needed such as
opening steam turbine inlet valves and or allowing some bypass flows.
5. Safety Features
Mk1 PB-FHR reactor protection functions are performed by a combination of passive
mechanisms, control signals and actuators from the safety-grade digital reactor protection
system (RPS), and from manual operator actions.
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 fueltemperature 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 boilingtemperature of flibe is 1430°C—as in light water reactors (LWRs), and larger than thegraphite moderator temperature reactivity feedback. Under the beyond design basis ATWS
accident where reactor scram does not occur upon loss of flow or loss of heat sink, the FHR
coolant equilibrates to a temperature close to the original fuel temperature. Simplified
analysis for the Mk1 design indicates that this equilibrium 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 naturalcirculation-driven heat transport 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 from the TCHXs is regulated by fail-open
valves that supply water to the thermosiphons integrated into these heat exchangers. The
valves are held closed during normal operation, and can also be closed to control overcooling during prolonged reactor shutdown. In addition to the passive emergency decay heatremoval provided by the DRACS, the PB-FHR power conversion system and the normal
shutdown cooling system provide heat removal capability and defense 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, byproviding natural-circulation heat removal and preventing chemical attack to fuel pebbles
from exposure to air. The PB-FHR utilizes a pool-type reactor configuration, similar to thedesign adapted for many sodium fast reactors. For BDBEs where the vessel leaks or ruptures,the Mk1 refractory cavity liner insulation system controls the level change in the vessel andprevents uncovering of fuel.
381

Containment system
The Mk1 design introduces another novel feature, a “gas gap” system, to make it physically
impossible to transmit excessive pressures to the reactor vessel and reactor
cavity/containment from potential tube or manifold pipe ruptures in a CTAH. The gas gap iscreated 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 thereactor cavity liner cooling system. Because these water pools also provide a source of waterfor evaporative cooling under beyond-design-basis event (BDBE) conditions, they areprovided 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, foreach 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 is scarce. The
advantage of the NACC system arises from additional revenues earned by providing flexible
grid support 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 the control system then arise from the economic
incentives to maximize system performance and to preserve the invested capital, thus highquality 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-change transients and turbine inlet temperature control (by bypassing
air around the CTAHs) for slower transients. Pump speed control is then used to control thecore temperature difference, and control rod position is used to control the averagetemperature. 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%.
The PB-FHR Post-Event Instrumentation (PEI) system will monitor primary systemtemperatures, as well as other safety related parameters, after design basis events. Data
provided by the PEI allows operators to assess the evolving plant state and take appropriate
long-term control actions; including assurance that overcooling does not occur.
The PB-FHR control system is designed to allow the reactor to withstand grid transients,
including loss-of-load, by venting hot air downstream of the LP CTAH, followed by acontrolled reactor power run-back and activation of the shutdown cooling system to remove
decay heat.

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The reactor control portion of the PB-FHR plant control system will have common sensor
elements to the RPS. All communications from the RPS to the control system will be via
buffered one-way communication links (i.e., IEEE 7.4.3.2-2003 Annex E type links). The
common sensing elements between the reactor safety and control system will include power
range neutron flux, primary coolant temperature, primary coolant level and inventory, andprimary coolant flow rate.
8. Plant Arrangement
Figure 4 presents a notional 180-acre site arrangement for a 12-unit Mk1 PB-FHR power
plant capable of producing 1,200 MWe base load and 2,900 MWe peak power output. Due tothe much smaller cooling requirements they do not need to be sited near bodies of water.
Population centers tend to be located near bodies of water which means that FHRs can besited in areas where fewer people want to live. So, rather than attempt to minimize the site
footprint, the more important goal is likely to facilitate construction of modules adjacent tooperating modules, and to optimize the degree to which some services are shared.

Figure 4: Mk1 site arrangement for a 12-unit, 180-acre PB-FHR plant, capable ofproducing 1200 MWe base load and 2900 MWe peak power(Reproduced courtesy of UC Berkeley)
Reactor building
The Mk1 reactor building and NACC system arrangements supports a multi-module plantconfiguration, shown in Figure 5, by allowing multiple units to be lined up in a row with aclear boundary between the reactor and its vital areas, versus the balance of plant (BOP). The
GT and associated equipment are configured to minimize the air pressure loss and circulating
power in the air ducting while maintaining a clear boundary between the reactors and the
BOP. This configuration makes it easier to co-locate combined nuclear services on one sideof 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 other side (off-site transmission, process steam loads and/or steam bottoming
turbines, cooling towers, etc.).

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The Mk1 reactor building is partially embedded below grade, with the reactor deck locatedslightly 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 steelplate/concrete composite (SC) modules, quite similar to the Westinghouse AP1000 shieldbuilding. The overall height and diameter of the Mk1 shield building are 47.5 m and 24.5 m,
respectively, compared to 83 m and 42 m for the 1150-MWe AP1000, so the Mk1 shieldbuilding volume is 2.2 times greater than the AP1000, per MWe baseload.

Figure 5: The Mk1 PB-FHR reactor shield building adjacent to the power conversion system(Reproduced courtesy of UC Berkeley)
The cylindrical shield building geometry provides a stiffer, lighter structure than is possible
with a rectangular building geometry. The low-leakage containment boundary is provided by
the liner of the reactor cavity and its gas space. With the containment boundary located wellinside of the reactor building external walls, the primary function of the cylindrical structure
is to provide an external events shield. It also acts as an element of the secondary filteredconfinement volume used for control of beryllium and radioactive contamination. Because
the refueling deck area of the reactor building must be accessed periodically during power
operation for fuel transfer work associated with on-line refueling, the refueling deck isdesigned to provide effective biological shielding.
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. Plant Economics
An overall capital cost of $4,500-$5,093/kW base load has been estimated depending on thenumber of units, ranging from one to twelve, present at the power plant site. Since all
infrastructure required to also perform peak load is already included the specific capital isreduced to $1,870-2,133/kW for base load plus peaking. These numbers should be comparedto capacity weighted average capital costs of a NGCC and a conventional nuclear power
plant, rather than each separately. Similar to capital construction costs, O&M costs drop from
$81.05/MWh (¢8.1/kWh) to $39.82/MWh (¢3.98/kWh) as the number of units per site
increase from one to twelve.
384

ANNEX I
SMR Estimated Timeline of Deployment

385

386

0-50

51-100101-150

151-200

201-250251-300> 301

Reactor Designs

IMR
AHWR-300
VBER-300
GTHTR300
IRIS
DMS
GT-MHR
EM2
BREST-OD-300
SC-HTGR
Westinghouse SMR
FUJI
MHR-T
ThorCon
LFTR
mPower
SMR-160
PBMR-400
IMSR
Flexblue
CAP150
HTR-PM
MSTW
Mk1 PB-FHR
SmAHTR
ACP100SMART
MHR-100
SVBR100
ACPR50S
CAREM25
NuScaleKLT-40SHTMR-100
G4M

•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•
•

ANNEX II
Summary of SMR Designs Based on Power Range

Power Range MW(e)Temperature (oC)

0

100

200

300

400

500

600

700

800

900

1000

0

5

10

20

25

30

System Temperature of
Water-cooled SMRs

35

Small Modular Reactor Design

15

System Temperature of Liquid
Metal- and Sodium-cooled SMRs

System Temperature of
Molten-salt SMRs

System Temperature ofGas-cooled SMRs

40

45

50

ANNEX III
Summary of SMR Designs Based on Core Exit Temperature

387

388

100

300

400

600

Pulp & paper manufacture

500

700

Thermochemical hydrogen production

Petroleum refining

900

1000

1100

(oC)

1200

Coal gasification

Methane reforming hydrogen production

800

Blast furnace steel making

Heavy oil desulfurization

Methanol production

Seawater desalination

District heating

200

Water cooled reactors

Liquid metal cooled reactors

Sodium-cooled fast reactors

Supercritical water-cooled reactors

Molten Salt reactors

Gas-cooled fast reactors

Very high temperature reactors

ANNEX IV
Summary of SMR Designs for Non-Electric Applications

ANNEX V
Bibliography
•
•
•
•

•

•
•
•
•

•

Contained status, design description and main features of
32 selected SMR designs;
Sorted by type/coolant: iPWR, PHWR, GCR, and LMFR;
Sorted by Country of Origin;
Included: CAREM (Argentina), FBNR (Brazil), CNP-300(China), Flexblue (France), IMR (Japan), SMART(Republic of Korea), ABV-6M (Russian Federation),
SHELF (Russian Federation), RITM-200 (RussianFederation), VK-300 (Russian Federation), 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), PFBR500 (India), BREST-OD-300 (Russian Federation),
SVBR-100 (Russian Federation), PRISM (USA), G4M(USA).
Published September 2012

Contained status, design description and main features 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), FBR-1&2(India), 4S (Japan), JSFR (Japan), PGSFR (Republic ofKorea), BN-1200 (Russian Federation), MBIR (RussianFederation), 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 2013

389

•
•
•
•

•

•
•
•

•

390

Contained status, design description and main features 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 (Republic ofKorea), ABV-6M (Russian Federation), SHELF (RussianFederation), RITM-200 (Russian Federation), VK-300(Russian Federation), VBER-300 (Russian Federation),
KLT-40S (Russian Federation), UNITHERM (RussianFederation), IRIS (International Consortium), mPower(USA), NuScale (USA), Westinghouse SMR(USA),SMR160 (USA), AHWR300-LEU (India), HTRPM (China), PBMR (South Africa), GT-MHR (RussianFederation),VVER-300 (Russian Federation), RUTA70(Russian Federation), ELENA (Russian Federation),
DMS (Japan), HTR-PM (China), GTHTR300 (Japan),
MHR-T (Russian Federation), MHR-100 (RussianFederation), 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), CPR1000 (China), HPR1000 (China), APR1400 (Republic ofKorea), APWR (Japan), AP1000(Japan), ABWR(Japan),
VVER1000(Russian Federation),, VVER1200(RussianFederation),.VVER1500(Russian Federation),
IPHWR(India), EPR(France), KERENA(France),
ATMEA1(France) , EC6 (Canada), ABWR(USA),
ESWR(USA)
Published September 2015

ANNEX VI
Acronyms
AC
ADS
ARIS
BCR
BDBA
BOP
BWR
CCWS
CEDM
CES
CMT
CPS
CRDM
CS
CSS
CTS
CV
CVCS
DAS
DBA
DC
DID
DLOFC
DVI
ECCS
ECDS
ECT
EDG
EHRS
EPZ
ESWS
FA
FE
FPU
FSAR
FSS
GDCS
GDWP
HEU
HHTS
HPCF
HTGR
HTR
HX
IC
I&C
LEU
LOCA
LOOP
LPFL

Alternating Current
Automatic Depressurization System
Advanced Reactor Information System
Back-up Control RoomBeyond Design Basis Accident
Balance of Plant
Boiling Water Reactor
Component Cooling Water System
Control Element Drive Mechanism
Containment Enclosure StructureCore Make-up Tank
Control and Protection System
Control Rod Drive Mechanism
Containment Structure
Control Safety System
Chemical Technological Sector
Containment Vessel
Chemical and Volume Control System
Diverse Actuation System
Design Basis Accident
Direct Current
Defence in Depth
Depressurized Loss of Forced Cooling
Direct Vessel Injection
Emergency Core Cooling System
Emergency Cooling Down System
Emergency Cooldown Tank
Emergency Diesel Generator
Emergency Heat Removal System
Emergency Planning Zone
Essential Service Water System
Fuel Assembly
Fuel Element
Floating Power Unit
Final Safety Analysis ReportFree Surface Separation
Gravity Driven Cooling System
Gravity Driven Water PoolHigh Enriched Uranium
Hybrid Heat Transport System
High Pressure Core Flooder
High Temperature Gas-cooled Reactor
High Temperature Reactor
Heat Exchanger
Isolation Condenser
Instrumentation and Control
Low Enriched Uranium
Loss of Coolant Accident
Loss of Offsite Power
Low Pressure Core Flooder

391

LWR
MCR
MHT
MOX
MSA
MW(e)
MW(th)
NDHP
NPP
NSSS
OBE
OCP
OTTO
PC
PCT
PCU
PCV
PCCS
PLOFC
PLS
PMS
PORV
PRHRS
PSAR
PSIS
PWR
RCIC
RCP
RCS
RCSS
RFA
RHRS
RP
RPV
RV
SBO
SG
SSE
TC
TEG
TEU
TM
TRISO
UCO
UHS
WDS
WPu
WWER

392

Light Water Reactor
Main Control Room
Main Heat Transport
Mixed Oxide
Moisture Separator Reheater
Mega Watt electric
Mega Watt thermal
Nuclear District Heating Plant
Nuclear Power Plant
Nuclear Steam Supply System
Operating Basis Earthquake
Outside Containment Pool
Once Through Then Out
Primary Containment
Peak Cladding TemperaturePower Conversion Unit
Primary Containment Vessel
Passive Containment Cooling System
Pressurized Loss of Forced Cooling
Plant Control System
Protection and safety Monitoring System
Power-Operated Relieve Valve
Passive Residual Heat Removal System
Preliminary Safety Analysis Report
Passive Safety Injection System
Pressurized Water Reactor
Reactor Core Isolation Cooling
Reactor Coolant Pump
Reactor Coolant System
Reactivity Control and Shutdown System
Robust Fuel Assembly
Residual Heat Removal System
Reactor Plant
Reactor Pressure Vessel
Reactor Vessel
Station Black-Out
Steam Generator
Safe Shutdown Earthquake
Turbo Compressor
Thermoelectric Generator
Thermoelectric Unit
Turbo Machine
Triple Coated Isotropic
Uranium Oxy Carbide
Ultimate Heat Sink
Waste Disposal System
Weapon-Grade Plutonium
Water Moderated Power Reactor

spine = 20.5 mm, 80gm paper

Advances in Small Modular Reactor
Technology Developments
Advances in Small Modular Reactor Technology Developments

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: http://www.iaea.org

16-24301

Printed by IAEA in Austria
August 2016

A Supplement to:
IAEA Advanced Reactors Information System (ARIS)
2016 Edition