Introduction To ThorCon TMSR500 Technology and Passive Safety System

Introduction to ThorCon TMSR 500 Technology and Passive Safety System
Based on the 7 Weeks of Training
July 2nd 2021 – August 13th 2021

by Dr. Kun Chen, the Chief Nuclear Officer of ThorCon Group
PT ThorCon Power Indonesia
September, 2021

i|Introduction to ThorCon TMSR 500 Technology and Passive

Safety System

INTRODUCTION TO THORCON TMSR 500
TECHNOLOGY AND PASSIVE SAFETY SYSTEM

Content:
Prof. Kun Chen, Ph.D

Editor:
THORCON POWER INDONESIA

Cover Illustration
THORCON POWER INDONESIA

First Edition, September 2021

Publisher
PT THORCON POWER INDONESIA
World Trade Center 3
Podium Retail Floor 3 (Spaces)
Jl. Jenderal Sudirman Kav 29 – 31
Karet, Setiabudi, Jakarta Selatan
DKI Jakarta 12920 – Indonesia
www.thorconpower.id

All rights reserved
ISBN 978–623–98311–1–0

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COPYRIGHT NOTICE

All the materials contained within this document are the intellectual
properties of PT ThorCon Power Indonesia. This document may not
be reproduced, distributed, transmitted, displayed, published, or
broadcast without the prior, express written permission by PT
ThorCon Power Indonesia. You may not alter or remove any
copyright or other notice from copies of this document.

© 2021 PT ThorCon Power Indonesia

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DISCLAIMER

Considering that not all training participants agree and sign the Non-
Disclosure Agreement, the images and designs displayed are not
100% the same as the current designs. However, these designs have
almost the same characteristics so that they are relevant in terms of
safety systems.

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FOREWORD BY HAENDRA SUBEKTI

By Haendra Subekti

Director for Regulation of Nuclear Installation

BAPETEN

On behalf of the BAPETEN, I would like to thank PT ThorCon Power Indonesia
for letting us have the opportunity to follow the training on the ThorCon 500
Molten Salt Reactor by ThorCon Power Indonesia. The seven meetings of the
training are long enough, but it is weekly, so it is not too hard. I'm hoping that we
can benefit from this training, from the materials presented by the experts, from
the discussion, and the questions and answers. The video-sharing on YouTube
is very interesting because if we miss any information, we can always rewatch
the training. I believe the sharing of knowledge, best practices, and experiences
between the experts and participants will provide benefits to our regulation
development and give valuable insight in establishing licensing schemes.

In addition, I hope this workshop can be a great way to enhance communication
between designers, experts, and the regulator. I got some points that ThorCon
has a roadmap to develop the design, the test facility, and the demonstration
plan. These are some developments that are very good for establishing the first
kind of nuclear power plant, or Molten Salt NPP, commercially. It is a very good
point and an opportunity that we have the process of regulation development for
power reactors in the aspect of design, some licensing process, and some liability
aspects. That are some important thing that we have to think about before we
develop some NPP in Indonesia. We have already communicated with the
ministry of energy regarding the sequence of the licensing process.

I think that is all from me. We are happy and appreciate what ThorCon Power did
to hold this training, and I hope you have any chance to have some other event
in the future. Thank you, Bob, Lars, Dr. Kun Chen, Dr. Manu, and PT ThorCon
Power Indonesia.

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PREFACE BY KUN CHEN

By Prof. Kun Chen
Chief Nuclear Officer
ThorCon Group

First of all, I have to thank everybody for participating in these workshops
over the past seven weeks. It is happening for a long time but I truly enjoyed
the discussion with everybody. We had very good participants. The
discussion was very insightful and they also benefit ThorCon in terms of
thinking about the design for the improvement.
I hope we can keep this communication in the future. Hopefully, we can have
more ways of communication with BAPETEN, BATAN, UGM, ITB, and
everybody else in Indonesia. I have been very impressed, I have been very
much enjoying this communication with everybody. Hopefully, in the future, I
will be able to visit Indonesia and have a face-to-face conversation with
people.
I want to thank Bob for organizing this wonderful event and the wonderful job
of having all these people together and I do like the way you organized this
workshop. Also Tagor, Heddy, Nafan, Zahra, Youri, and everybody else in
ThorCon Indonesia, you have done a wonderful job. Thank you all.

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PREFACE BY BOB S. EFFENDI

By Bob S. Effendi
Chief Operating Officer
PT ThorCon Power Indonesia

Firstly, on behalf of PT ThorCon Power Indonesia, I would like to show our
appreciation to all 68 participants who participated in 7 weeks of training with a
total of 21 hours of ThorCon Technology and Passive Safety Systems conducted
by Prof. Dr. Kun Chen, the Chief Nuclear Officer of ThorCon, and a special
appreciation to Nuclear Energy Regulatory Agency of Indonesia (BAPETEN),
which sent 32 people to participate. We think this training will be greatly
beneficial, especially for BAPETEN to better understand ThorCon’s Thorium
Molten Salt Reactor design in preparation for the licensing process, which we
hope will start in the near future.
Secondly, this document, which we published as a public document, is the
training material presented in narrative and descriptive format with all the
references included, which we hope will be used as a reference for Indonesian
nuclear experts and students who want to understand the ThorCon design better
and to enrich the Indonesian nuclear community's knowledge of liquid-fuel
design. We will give this document to all institutions involved in this training,
namely BAPETEN, BATAN, UGM, and ITB.
Lastly, I would like to say thanks to all the ThorCon personnel involved in making
the 7-week training a success, especially my personal thanks to Tagor Malem
Sembiring and his assistant, Lee Youri M, for making and preparing this
document.

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TABLE OF CONTENTS

FOREWORD BY HAENDRA SUBEKTI ............................................................v
PREFACE BY KUN CHEN...............................................................................vi
PREFACE BY BOB S. EFFENDI .................................................................... vii
TABLE OF CONTENTS ................................................................................. viii
LIST OF FIGURES............................................................................................x
LIST OF TABLES .......................................................................................... xiii
Chapter 1. History and Current Status of Molten Salt Reactor.................... 1

1.1. What is MSR? ........................................................................................ 1
1.2. History of MSR in the U.S....................................................................... 1
1.3. Why MSR Stopped in the U.S. ? ............................................................ 4
1.4. The Current Status of MSRs in the World............................................... 6
1.5 The Taxonomy of MSRs .......................................................................... 6
1.6. Comparison of MSRs ........................................................................... 12
1.7. Molten Salt Reactor Research and Development in SINAP, China....... 12
1.8. Regulatory Process in China and Canada ............................................ 13
1.9. Some Critical Issues Related to MSR that are often Asked .................. 15
Chapter 2. TMSR 500 Plant and MSR Safety Features ............................... 20
2.1. Introduction of TMSR 500 Plant............................................................ 20
2.2. TMSR 500’s Safety Philosophy ............................................................ 32
2.3. Safety Features of Molten Salt ............................................................. 32

2.3.1 How to achieve control ? ................................................................. 33
2.3.2. How to achieve cool ? .................................................................... 34
2.3.3. How to Achieve Contain ? .............................................................. 37
Chapter 3. Neutronic and Thermal-Hydraulic of TMSR 500 Plant ............. 40
3.1. Current Status of Reactor Can Design of TMSR 500 Plant................... 40
3.2. Reactor Can Components .................................................................... 41
3.3. Neutronic and Thermal-Hydraulics Aspects.......................................... 49
3.4. Principles of Fuel Management ............................................................ 58
Chapter 4. Operation and Balance of Plant of TMSR 500 .......................... 62
4.1. Operation of TMSR 500........................................................................ 62
4.2. Load Following Power Operation.......................................................... 68
4.3. Balance of Plant (BOP) ........................................................................ 71

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Chapter 5. Off-gas, Fuel Transfer, and Instrumentation and Control
Systems of TMSR 500................................................................ 75

5.1. Fuel Transfer System ........................................................................... 75
5.2. Off-gas Systems................................................................................... 79
5.3. Instrumentation and Control System (I&C) ........................................... 91
Chapter 6. TMSR 500 Design and Safety Assessment Process and Typical

MSR Events................................................................................ 99
6.1. Safety Philosophy................................................................................. 99
6.2. Safety Objectives, Goals, and Acceptance Criteria............................. 100
6.3. Design Process .................................................................................. 101
6.4. Safety Assessment............................................................................. 105
6.5. Analysis of Scenario like Fukushima Daiichi ....................................... 122
6.6. Analysis of Other MSR Events ........................................................... 127
REFERENCES ............................................................................................. 130
INDEX ............................................................................................................133
LIST OF PARTICIPANTS ............................................................................ 135
PARTICIPANT REVIEWS ............................................................................ 137
The Autobiography of Prof. Kun Chen, Ph.D............................................ 138

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LIST OF FIGURES

Figure 1.1 Aircraft Reactor Experiment [Ergen et al., 1957] ........................... 2
Figure 1.2 The Reactor Core of MSRE [ORNL, 2015].................................... 3
Figure 1.3 U of California MK1 PB FHR [Andreades, 2016] ........................... 6
Figure 1.4 ORNL SmAHTR [Greene et al., 2010]........................................... 7
Figure 1.5 ORNL Single fluid MSBR [Robertson et al, 1971] ......................... 7
Figure 1.6 ORNL Two fluids MSBR [Briggs et al.,1964] ................................. 7
Figure 1.7 TMSR 500 [Jorgensen, 2017] ....................................................... 8
Figure 1.8 Japanese Fuji [Furukawa et al.,1987]............................................ 8
Figure 1.9 SINAP-LF1 [SINAP, 2021] ............................................................ 8
Figure 1.10 French MSFR [Mathieu et al., 2006] ............................................. 9
Figure 1.11 Russian MOSART Burner [Ignatiev et al., 2007] ........................... 9
Figure 1.12 SWISS SOFT [Taube et al., 1972] ................................................ 9
Figure 1.13 ORNL ARE [Fraas and Sovalainen, 1954] .................................. 10
Figure 1.14 Seaborg CMSR [Schonfeld et al., 2018]...................................... 10
Figure 1.15 Moltex SSR-W [Scott et al., 2015] ............................................... 10
Figure 1.16 German DFR [Huke et al., 2015] ................................................. 11
Figure 2.1 2 × 500 MW TMSR 500 Power Plant .......................................... 20
Figure 2.2 Basic Shipyard Terms ................................................................. 21
Figure 2.3 TMSR 500 Overview................................................................... 22
Figure 2.4 TMSR 500 Cutaway from Starboard Side ................................... 23
Figure 2.5 TMSR 500 Cutaway from Port Side ............................................ 24
Figure 2.6 Plan View of Turbine Hall at TG Deck Level................................ 25
Figure 2.7 Plan View of Turbine Hall from Starboard Side ........................... 25
Figure 2.8 Plan View of Turbine Hall from Port Side .................................... 25
Figure 2.9 Plan View of Turbine Hall from Aft Side ...................................... 26
Figure 2.10 Plan View of Condenser from Forward Side ............................... 26
Figure 2.11 Plan View of GIS Hall from Starboard Side ................................. 27
Figure 2.12 Plan View of GIS Hall.................................................................. 27
Figure 2.13 TMSR 500 Sections .................................................................... 29
Figure 2.14 Energy Flow Diagram ................................................................. 31
Figure 2.15 Freeze Valve and Drain Tank ..................................................... 34
Figure 2.16 The Sentry Turbine of TMSR 500 ............................................... 35

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Figure 2.17 Coldwall and Cooling Pond ......................................................... 36
Figure 2.18 Basement Water and Ballast Tank .............................................. 37
Figure 2.19 Multiple Steel Barriers ................................................................. 38
Figure 3.1. Comparison of the Old and Current Can Design of TMSR 500

[IAEA2020] .................................................................................. 40
Figure 3.2 Inside the Can............................................................................. 41
Figure 3.3. The old design of TMSR 500 Core and Fuel Log ........................ 43
Figure 3.4. Primary Heat Exchanger of TMSR 500 ...................................... 47
Figure 3.5 Multiple Freeze Valve ................................................................. 48
Figure 3.6 The Reactivity of the Core during the Draining............................ 51
Figure 3.7 Drain Tank Creep Life Depletion in 1 Drain ................................. 51
Figure 3.8 Sketch View of Drain Tank. ......................................................... 52
Figure 3.9 Effective Equivalent Xe-135 in Salt ............................................. 54
Figure 3.10 Fast Effective Removal Rate....................................................... 55
Figure 3.11 Decay Heat after 15 Days Operation at Nominal Power.............. 57
Figure 3.12 Principles of Fuel Management for One Power Module .............. 58
Figure 4.1 Review of Plant States ................................................................ 62
Figure 4.2 Review of Fission Island ............................................................. 63
Figure 4.3 Normal Operating Modes ............................................................ 64
Figure 4.4 Load Following System ............................................................... 69
Figure 4.5 The Steam Cycle Temperature Curve......................................... 70
Figure 4.6 The Tertiary Salt Temperature Curve.......................................... 70
Figure 4.7 The Secondary Salt Temperature Curve ..................................... 70
Figure 4.8 The Fuelsalt Temperature Curve ................................................ 70
Figure 4.9 Pump Speed and Bypass Curve ................................................. 70
Figure 4.10 Diagram of the Steam Cycle ....................................................... 73
Figure 5.1 Simplified Fuel Transfer Diagram................................................ 75
Figure 5.2 Plan View of the Can Components ............................................. 76
Figure 5.3 Cross Section of the FDT............................................................ 77
Figure 5.4 Plan View of the FDT Components ............................................. 77
Figure 5.5 Fuelsalt Lines Out of the FDT ..................................................... 78
Figure 5.6 Off-gas System Process ............................................................. 81
Figure 5.7 Off-gas System Process by Functions in Normal Operation ........ 83
Figure 5.8 Flow Diagram of Each Modes ..................................................... 84
Figure 5.9 Off-gas Recuperators.................................................................. 85

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Figure 5.10 The Cold Traps ........................................................................... 86
Figure 5.11 Location of the off-gas components ............................................ 87
Figure 5.12 Off-gas Flow and Containment Diagram ..................................... 88
Figure 5.13 Review of Plant Items ................................................................. 91
Figure 5.14 The Simplified I&C Architecture .................................................. 92
Figure 5.15 Main Control Room Location....................................................... 93
Figure 5.16 Backup Control Room and Control Systems Room..................... 94
Figure 5.17 Backup Control and the Control System Location ....................... 95
Figure 6.1 TMSR 500 Design Process Overview – Conceptual Engineering

Design ....................................................................................... 102
Figure 6.2 TMSR 500 Design Process Overview – Preliminary Engineering

Design ....................................................................................... 103
Figure 6.3 TMSR 500 Design Process Overview – Detail Engineering Design

.................................................................................................. 104
Figure 6.4 TMSR 500 Design Process Overview – Detail Design Validation

.................................................................................................. 105
Figure 6.5 Overview of the Safety Assessment Process ............................ 106
Figure 6.6 Curve of the Fuelsalt Volume Change Over the Time ............... 124
Figure 6.7 Curve of the Fuelsalt Flow Rate Change Over the Time ........... 124
Figure 6.8 Fuelsalt Temperatures and Flow Rates at the Freeze Valve ..... 125
Figure 6.9 Drain Tank Heat Transfer Transient .......................................... 126
Figure 6.10 Drain Tank Creep Life Depletion In 1 Drain............................... 126

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LIST OF TABLES

Table 1.1 The Taxonomy of MSRs ................................................................... 6
Table 1.2. Difference of the ORNL-MSRE, SINAP-TMSR-LF1, and ThorCon-

TMSR 500 ................................................................................... 12
Table 1.3 Licensing Process in China ............................................................ 14
Table 1.4 Licensing process in Canada.......................................................... 15
Table 2.1 Part of the Shipyard Elaboration and Illustration ............................. 25
Table 3.1 Characteristic of the Can ................................................................ 42
Table 4.1 Summary of the Detailed Modes..................................................... 67
Table 5.1 Fuel Transfer Components ............................................................. 76
Table 5.2 Decay of Kr and Xe Isotopes .......................................................... 89
Table 5.3 Isotope Accumulation in Off-gas System ........................................ 90
Table 6.1 Reactor PIEs Group ..................................................................... 109
Table 6.2 Safety System PIEs Group ........................................................... 114
Table 6.3 Power Conversion PIEs Group ..................................................... 115
Table 6.4 Plant Infrastructure PIEs Group .................................................... 116
Table 6.5 External Hazards PIEs Group....................................................... 117
Table 6.6 Fuelsalt and Can Transfer PIEs.................................................... 119
Table 6.7 Operator Action PIEs Group ......................................................... 120
Table 6.8 Historic PIEs Group ...................................................................... 121
Table 6.9 Sequence of Fukushima Events ................................................... 123

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History and Current Status of Molten Salt Reactor

Chapter 1. History and Current Status of Molten Salt
Reactor

1.1. What is MSR?

The Molten Salt Reactor (MSR) has a broad and diverse category of nuclear
reactors in which a molten salt plays one or more significant functions in the
reactor core, such as fuel chemical form, fuel dissolved in liquid salt, coolant or
moderator [Worrall et al., 2018]. If the reactor uses molten salt for fuel, coolant,
or moderator, it can be considered as one of MSR. By definition, there are many
reactor designs in the MSR family.
Molten Salt Reactor is one of the six Generation IV nuclear reactors recognized
by the Generation IV International Forum (GIF). According to the GIF, the
Generation IV reactors “will use fuel more efficiently, reduce waste production,
be economically competitive, and meet stringent standards of safety and
proliferation resistance.”
1.2. History of MSR in the U.S.

Molten Salt Reactor was developed at Oak Ridge National Laboratory (ORNL)
after World War II for the military aircraft power. The plan is to use an aircraft
propulsion reactor to elevate the plane through the air. Figure 1.1 depicts the
world's first MSR, the Aircraft Reactor Experiment (ARE) in 1954. The ARE was
used to study the nuclear stability of the circulating fuel system and the use of
molten fluoride fuels for aircraft propulsion reactors [Worrall et al., 2018]. The
fuelsalt of the ARE was a mixture of NaF-ZrF4-UF4, the moderator was BeO, and
all of the piping was Inconel. The enrichment percentage of U-235 was 93%. It
ran for nine days at steady-state outlet temperatures up to 1580°F (1133 K) and
powers of up to 2.5 MWth. There were no mechanical or chemical issues, and
the reactor was found to be stable and self-regulating [Worrall et al., 2018].

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History and Current Status of Molten Salt Reactor

Figure 1.1 Aircraft Reactor Experiment [Ergen et al., 1957]
The success of ARE resulted in the most famous MSR in the world, the Molten
Salt Reactor Experiment (MSRE). On June 1, 1965, MSRE reached criticality for
the first time. It ran at full power for nearly 8 MWth (7.34 MWth) for 13271 hours
(1.5 years). The MSRE fuelsalt was LiF-BeF2-ZrF4-UF4, and graphite served as
the moderator. The MSRE provided a lot of the available information regarding
MSR, and almost all designs nowadays are related to the MSRE. The MSRE ran
on U-235/U-238 or U-233 fuels. The intention of using U-233 in the MSRE was
to prepare for future thorium fuel. The MSRE was developed as part of the
thorium/uranium fuel cycle.

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History and Current Status of Molten Salt Reactor

Figure 1.2 The Reactor Core of MSRE [ORNL, 2015]
The MSRE is very famous, and at the same time, there is a lot of MSR
development going on in other countries, such as:

1. Germany. The concept, called Molten Salt Epithermal (MOSEL), was
proposed in the 1960s.

2. The United Kingdom. The salt chemistry work started in 1965.
3. China. The MSR for thorium started at the Shanghai Institute of Applied

Physics (SINAP) in 1970.
4. Netherlands. The Technische Universiteit (TU) Delft developed the MSR

components in the 1970s.
5. Switzerland. The concept called SOFT was proposed by the Paul

Scherrer Institute (PSI) in the late 1970s.

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6. Russia. The MSR program was started at the Kurchatov Institute in 1976.
7. France. Electricite de France (EDF) proposed the Actinide Molten Salt

Transmutter (AMSTER) concept, and the Thorium Molten Salt Reactor
(TMSR) was proposed by the Centre National de la Recherche
Scientifique (CNRS) in the late 1990s.

1.3. Why MSR Stopped in the U.S. ?
Molten Salt Reactor Experiment was a massive success because it ran for four
years and met all of its objectives. The natural step (TNS) in the development is
to build a larger high-power demonstration reactor to accomplish and enable the
technology's commercialization. In 1972, the United States Atomic Energy
Commission (AEC), the sole government agency that oversees of all nuclear
activities in the United States, assessed whether a large demonstration plant
should be built in the United States. The conclusion and recommendations of the
evaluation for the next steps were documented in the WASH-1222 report.

According to the WASH-1222 report, MSR was not ranked high enough on the
US government's list of priorities for energy development. At the time, the Sodium
Fast Reactor (SFR) was ranked higher than all other technologies by the United
States government. The government recognized that it lacked the necessary
resources (money) to support the parallel development of all technologies.
Therefore, they dismissed MSR, the High-Temperature Gas-Cooled Reactor
(HTGR), and several other technologies.

The WASH-1222 also identified several technical issues that require further
study:

1. The neutron embrittlement of nickel-based alloys at high temperatures.
The MSRE uses a nickel-based alloy called Hastelloy-N and there are
several other versions of that. If the neutron fluence was too high or the
vessel was put into service for too long, the nickel-based alloy would
have an embrittlement effect.

2. The radiation damage and dimensional change to graphite at high fast
neutron fluences. Graphite has a particular characteristic. It will expand
with neutron irradiation and then shrink after reaching certain
displacement-per-atoms (DPAs).

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History and Current Status of Molten Salt Reactor

3. The liquid-liquid chemical extraction process, if fuel reprocessing is
desired. Because the ORNL was going to reprocess the fuel, a chemical
process had to be developed.

4. The tritium permeation of the alloy at high temperatures. The MSRE uses
lithium, which generates a relatively large amount of tritium. At high
temperatures, the tritium can penetrate the alloy. This issue needs to be
resolved.

5. The development of significant components. The MSRE used a special
nickel-based alloy that is relatively small because it only has 8 MW.
However, the more substantial power density of the power plant will need
a more prominent nickel-based alloy component which was not available
at that time.

6. The understanding of fission products. It is about the release of cesium
and iodine, and other essential elements.

7. Remote inspection and maintenance. If the power plant were planned to
be operated for 30 or 40 years, it would need review and maintenance.
Therefore, the development of a robotic device is required.

Because of the AEC's observations following the evaluation, the MSR program
in the United States was refocused for the next ten years after the 1970s on
resolving the technical issues identified in this report. The technical problems
were resolved mainly by the end of the 1970s, and the United States was ready
to develop a large demonstration MSR, but the opportunity had passed. Until
recently, the MSR was not considered significant enough to warrant the
necessary resources for the next development phase in the United States. By
the early 1980s, several nuclear accidents had occurred, and public opinion in
the United States had shifted, so the MSR was never given adequate resources
to be developed for commercial use.

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History and Current Status of Molten Salt Reactor

1.4. The Current Status of MSRs in the World
From the 1960s to the 1970s, MSR was a hot topic. It vanishes in the 1980s but
has recently been revived. Numerous government and private sector efforts have
resurfaced in the last ten years for various reasons, including climate change and
low carbon emissions. Now there are multiple MSR concepts and options
available around the world based on features such as:

1. Fuel: uranium, thorium, or plutonium;
2. Neutron spectrum: thermal or fast;
3. Conversion ratio: burner, breeder, or converter;
4. Salt: fluoride or chloride;
5. Core type: single-fluid, two-fluid, or no-fluid;
6. Reactor site: land-based or non-land-based, floating or non-floating.
The IAEA is creating a taxonomy of all available MSR concepts to make them
comprehensible.

1.5 The Taxonomy of MSRs
According to IAEA, four major MSR classes can be seen in Table 1.1.
Table 1.1 The Taxonomy of MSRs

Type Salt
1.1 Cooled
Reactor
Class I Family 1 with
(Fluoride Pebble
Bed Fuel
Salt
Cooled
MSRs)

Figure 1.3 U of California MK1 PB FHR
[Andreades, 2016]

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Type Salt
1.2 Cooled
Reactor
with Fixed
Fuel

Type Figure 1.4 ORNL SmAHTR [Greene et al.,
2.1 2010]

Single-fluid
Th-U
breeder

Family 2 Figure 1.5 ORNL Single fluid MSBR
(Graphite [Robertson et al, 1971]
Moderate
d MSRs) Type Two-fluids
2.2 Th-U
breeder

Figure 1.6 ORNL Two fluids MSBR [Briggs et
al.,1964]

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Type Uranium
2.3 Converters
and Other
Concepts

Figure 1.7 TMSR 500 [Jorgensen, 2017]

Figure 1.8 Japanese Fuji [Furukawa et
al.,1987]

Figure 1.9 SINAP-LF1 [SINAP, 2021]

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Type Fluoride
3.1 Fast Th-U
Breeder

Family 3 Figure 1.10 French MSFR [Mathieu et al.,
(Homo- 2006]
geneous
Fluoride Type Pu-fuelled
3.2 Fluoride
Fast Fast
MSRs) Reactor

Class
II

Figure 1.11 Russian MOSART Burner
[Ignatiev et al., 2007]

Type Chloride
4.1 Fast
Breeder

Family 4 Type Chloride
(Homo- 4.2 Fast Breed
geneous and Burn
Chloride Reactor

Fast
MSRs)

Figure 1.12 SWISS SOFT [Taube et al.,
1972]

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Type Solid
5.1 Moderator
Heteroge-
neus
Reactor

Family 5 Figure 1.13 ORNL ARE [Fraas and
(Non- Sovalainen, 1954]

graphite Type Liquid
Moderate 5.2 Moderator
d MSRs) Heteroge-
neus
Class Reactor
III

Figure 1.14 Seaborg CMSR [Schonfeld et al.,
2018]

Type Heteroge-
6.1 neus Salt
Cooled
Family 6 Fast
(Hetero- Reactor
geneous
Chloride

Fast
MSRs)

Figure 1.15 Moltex SSR-W [Scott et al., 2015]

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Type Heteroge-
6.2 neus Lead
Cooled
Fast
Reactor

Figure 1.16 German DFR [Huke et al., 2015]

Class • Directly Cooled MSRs
IV • Subcritical MSRs
• Hybrid Moderator MSRs
• Chloride Salt Cooled Fast Reactors
• Frozen Salt Reactors
• Hybrid Spectrum MSRs
• Heterogeneous Gas Cooled MSRs

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1.6. Comparison of MSRs
The ORNL-MSRE, SINAP-TMSR-LF1, and ThorCon-TMSR 500 belong to the
same type and thus are closely related. As an outcome, it is worthwhile to
compare those three to see how they differ from the common features. Table 1.2
presents a comparison of those MSRs.

Table 1.2. Difference of the ORNL-MSRE, SINAP-TMSR-LF1, and ThorCon-
TMSR 500

Features ORNL-MSRE SINAP-TMSR- ThorCon-TMSR 500
LF1
Power 7.34 MWth 250 MWe × 2
Fuel U-235, U-233, Pu-239 2 MWth U-235, Th-232
Fuelsalt LiF, BeF2, ZrF4 U-235, Th-232 NaF, BeF2, ZrF4
Moderator Graphite LiF, BeF2, ZrF4 Graphite
Structure Material Nickel-based Alloy Graphite Stainless Steel 316
Nickel-based
Shutdown Rods Alloy Rods
Rods Drain Tank, Cold
Passive Heat Drain Tank, Water Cold Wall Wall
Removal Tube

The ThorCon-TMSR 500 differs from the ORNL-MSRE and SINAP-TMSR-LF1
in terms of mitigating some issues. TMSR 500 replaces LiF with NaF to prevent
the generation of much tritium during reactor operation. In addition, TMSR 500
uses stainless steel over nickel-based alloys because the nickel-based alloys are
subject to neutron embrittlement, and also, stainless steel is easier to procure
and manufacture.

1.7. Molten Salt Reactor Research and Development in SINAP, China
The MSR project in China started in 2011. The government was concerned about
uranium supply because China has many Pressurized Water Reactors (PWRs)
that are operational or under construction, but uranium is primarily supplied by
Australia and Kazakhstan. Therefore, the government is looking for thorium to
have some domestic fuel supplies. They are willing to invest up to USD 500
million to MSR Development because the MSR has the potential to use Chinese
thorium resources. The government also wants to investigate nuclear heat
application, which has been discussed in the nuclear industry for over 20 years
but has yet to be commercialized.

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The development of MSR technology requires the construction of two types of
equipment or facilities:

1. Laboratory access to molten salt chemistry, corrosion, and material
study; and

2. Molten salt loops for thermal-hydraulics and operational experience.

Material components are classified into three types: alloy for structural materials,
salt for coolant and fuel, and graphite for the moderator. SINAP has contracted
with several suppliers to produce the alloys, salt, and graphite.

Enrichment of Li-7 was successful in the laboratory, but not so much on a larger
scale. SINAP or any of their suppliers had not yet achieved a significant amount
of Li-7 enrichment. The Li-7 that SINAP is currently using for testing was
purchased from Russia, which uses mercury technology. As a result of health
concerns, mercury technology has been banned in many countries around the
world. As a consequence, the Li-7 continues to be a barrier to widespread MSR
deployment.

A pre-fission testbed platform known as TMSR-SFO was developed by SINAP.
The plan is to design and construct a facility similar to the reactor, but instead of
using nuclear fuel as a heat source, they will use electric heaters. The testbed
was built to conduct some experiments as well as gain experience in the
fabrication of the components and the operation of the facility. When the facility
is operational for the first time and some problems arise, they will be much easier
to resolve than at the nuclear power plant. SINAP will scale up to a power plant
called TMSR-LF1 once the pre-fission testbed platform is completed.

1.8. Regulatory Process in China and Canada
a. China
In China, the regulatory framework has several layers, such as:

1. Nuclear Safety Law
2. Nuclear Safety Regulations (HAF), issued by the State Council or

ministry
3. Nuclear Safety Guides (HAD), issued by the ministry
4. Mandatory National Standards (GB)
5. Recommended National Standards (GB/T)

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The regulatory body in China is the Ministry of Ecology and Environment and the
National Nuclear Safety Administration, which is solely responsible for regulating
the nuclear part.

The licensing process in China shown in the Table 1.3.

License for: Table 1.3 Licensing Process in China
Site
Requirements
Site SAR, EIA, etc.

Construction PSAR, EIA, QA, etc. (valid for 10 years)

Initial fuel loading FSAR, QA, ER, etc.

Operation Maintenance plan, In-service Inspection plan, refueling plan, pre-
service inspection report, initial fuel loading report, etc. (valid for
design lifetime)

Decommissioning Decommissioning SAR, EIA, QA, etc.

EIA = Environmental Impact Assessment
QA = Quality Assurance
SAR = Safety Analysis Report
PSAR = Preliminary Safety Analysis Report
FSAR = Final Safety Analysis Report

The regulatory in China also offers optional processes or services. For example,
they provided to combine licenses permitted (e.g., factory assembled floating
plants, fuel cycle plants, small test facilities, etc.) and pre-licensing engagement
encouraged.

b. Canada
The regulatory body in Canada is the Canadian Nuclear Safety Commission
(CNSC), and the regulatory framework in Canada are:

1. Act passed by Parliament, e.g., NSCA;
2. Regulations, made by the Canadian Nuclear Safety Commission

(CNSC) authorized under the NSCA and approved by the Governor in
Council;
3. Licenses and certificates, issued by the CNSC;
4. Regulatory document (REGDOC), issued by the CNSC.
After 2019, there is a new act passed by the parliament. It is called the impact
assessment act. It expands the requirements for environmental impact
assessment to include social-economic assessment.
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The licensing process in Canada shown in the Table 1.4.

Table 1.4 Licensing process in Canada

License to: Duration (months) Requirement

Prepare site 24

Construct 32
Operate Impact Assessment

24

Decommission 24

The CNSC offers optional processes or services. The first offer is to combine
licenses to construct and operate to reduce the time (40 months). The second
offer is pre-licensing engagements such as the Vendor Design Review (VDR)
and the 4-step process to determine the licensing strategy.

1.9. Some Critical Issues Related to MSR that are often Asked

a. Corrosion issues of using sodium fluoride and other issues associated with the
use of sodium compared to the use of lithium

Lithium fluoride has the benefit of the graphite moderator if compared with
sodium fluoride. The element of lithium is lighter than sodium, so it has better
moderation. Several vendors, companies, and laboratories are not using lithium
because it is hard to get the enriched Li-7. Natural lithium has a large amount of
Li-6, which under neutron irradiation will generate tritium. Lithium should be
enriched to comprise more than 99.9% Li-7. Unfortunately, it is difficult the enrich
lithium. The only available method of enrichment is by using Mercury, but most
countries, including the United States, Europe, China, and Japan, have all
banned the use of Mercury technology.

Due to Li-7 enrichment issues, its capability is insufficient for large-scale MSR
deployment. To address this issue, MSR developers are switching to a different
salt, such as sodium. Switching to sodium solves the "you do not have enough
supply" issue, but it has less moderation. In addition, due to the less-than-ideal
neutron performance, the fuel use, physical material use, and negative reactivity
coefficient will be less efficient.

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In terms of resistance to corrosion, sodium and lithium are nearly identical. The
chemistry of salt control is always necessary for any salt and any structured
material. The oxygen level should be kept below a few hundred parts per million
(ppm). In comparison to nickel, stainless steel will necessitate stringent salt
chemistry control, thicker vessel walls, and/or reduce the operating time by
replacing components regularly to resolve the corrosion issue.

b. What are the main technical challenges for MSR development nowadays? Is
it finding the best salt composition, or is it finding the best reactor
configuration? Or is it understanding how the material behaves under salt
irradiation or integrating the re-processing to the MSR itself?

The materials are the most complex technical challenge in nuclear technology.
The most challenging aspect of working with molten salt is the structure of the
material. MSR developers experimented with a nickel-based alloy and stainless
steel. However, the nickel-based alloy is both expensive and challenging to
produce. Stainless steel is less costly and easier to manufacture, but it is less
resistant to corrosion under the same circumstances as the nickel-based alloy.

Corrosion, neutron embrittlement, and other challenges are considered to be
manageable. But because there has not been a molten salt reactor built since
the 1960s, people do not have experience yet. There are a lot of new
technologies that can be used but have not been tested. The developer has to
be allowed to test those ideas and gain the data to resolve the molten salt issues.

In terms of the graphite moderator, all MSRs have to use high-quality graphite.
MSR graphite should have relatively small pores and cannot be porous so that
salt does not penetrate too thoroughly. Other than the Li-7, the most challenging
thing in salt production is the production of high-quality salt. Several kilos of high-
quality salt can be produced, but for 100 tons, the chemical process for producing
high-quality salt must be enhanced. Other challenges, such as high
temperatures, necessitate the use of a special instrument, special operating
procedures, and special components, but these can all be overcome.

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c. How does TMSR 500 address the technical challenges of MSR?

In terms of material, the TMSR 500 is made of stainless steel, which eradicates
the need for nickel-based alloy treatment. As a result, stainless steel is a good
candidate in that particular respect. To prevent corrosion, TMSR 500 will have to
control the chemistry of the salt. The advantage of the TMSR 500's four to eight-
year design lifetime is that the components can be replaced regularly to resolve
the corrosion issue. TMSR 500 uses sodium salt instead of lithium salt, so there
is no lithium issue. In terms of salt, the challenge is to find suppliers and help
them to scale up their capacity to make the salt.

Currently, ThorCon is ready to address those challenges by collaborating with
the Bandung Institute of Technology (ITB), the University of Wisconsin, and
Virginia Technology to do research and development of the molten fuel salt in
a laboratory in Jatinangor, Bandung, Indonesia.

For graphite, there are some challenges. But the benefit of the TMSR 500 is that
the graphite will only be used for four years. The replacement time was
determined by the lifetime of the graphite. The lifetime of the components cannot
be confirmed if they can last for 80 years because many phenomena can happen
in 80 years. Therefore, the strategy is to replace the components every four
years. It is within the knowledge boundary because it was the time that MSRE
operated.

d. How can it be claimed that TMSR 500 is the scale-up of the MSRE?

The structure of the reactor is very similar to MSRE, but there is a replacement
of elements in that structure. For example, replacing Li-7 salts with sodium has
changed several aspects for a good reason, but the overall design is very similar.

e. How has the experience of regulatory in China to review the SINAP reactor?

There were general requirements for radiation protection and quality assurance
for reactors, but there were no special requirements for molten salt because there
was no molten salt licensed in China in the past. The National Nuclear Security
Agency (NNSA) did not issue any special requirements for SINAP; instead,
NNSA asked SINAP to develop a set of requirements that NNSA would review.

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As a new technology, SINAP may have better knowledge than the NNSA in terms
of the molten salt reactor.
The NNSA then did one thing. They summarized the design requirements
discussion with SINAP and established a review guide for their reviewers.
Because SINAP and Prof. Kun was familiar with the technology, NNSA was able
to use the applicant as a resource to develop those requirements and guidance.
They reviewed it, agreed on it, and gained technology as an outcome of the
discussion.

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SUMMARIES

1. Molten Salt Reactors (MSR) are a broad and diverse category of reactors,
in which molten salt plays one or more significant functions in the reactor
core, such as fuel chemical form, liquid fuel carrier, coolant, or moderator.

2. MSRE was the first MSR in the world that operating for four years. Most
of the information about MSR at the time came from the MSRE, and
almost all of the designs today are in some way related to the MSRE.

3. There are many MSR concepts around the world today that are based on
the fuel, neutron spectrum, chemical salt, sources, types of cores, and
their locations.

4. The TMSR 500 is a scaled-up MSRE. The reactor structure is very similar
to MSRE, but some replacement elements have been added to improve
the performance of the reactor.

5. The LiF is replaced by the NaF in the TMSR 500. The Li-7 in LiF must be
enriched, and there will still be a lot of tritium produced by neutron
irradiation.

6. Stainless steel is used instead of a nickel-based alloy in the TMSR 500.
The reason is that nickel-based alloy is expensive and difficult to obtain
and manufacture, but stainless steel components are not.
YOUTUBE VIDEO OF THE TRAINING RECORD:

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Chapter 2. TMSR 500 Plant and MSR Safety Features

2.1. Introduction of TMSR 500 Plant
An artistic visualization of two TMSR 500 is shown in Figure 2.1. There are two
power plants, each with a capacity of 500 MWe. In the middle of the plant, there
is a ship called the Canship that transports the fuel and core to the power plant.
Figure 2.1 shows an open vault on the right. The crane will use this vault to lift
the core from the power plant and place it on the Canship. That is how to switch
the core. The first demonstration reactor will only have one 500 MWe power
plant.

Figure 2.1 2 × 500 MW TMSR 500 Power Plant
The layout of the TMSR 500 is quite different and unique when compared to the
traditional nuclear power plant layout, which is generally flat. The TMSR 500
consolidates everything into a single large structure that resembles a ship. As
shown in Figure 2.2, TMSR 500 describes the power plant using shipyard
terminology.

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Figure 2.2 Basic Shipyard Terms

The terminology is listed below:

1. Bow : the front of the plant

2. Stern : the rear of the plant

3. Forward : the direction towards the bow

4. Aft : the direction towards the stern

5. Port Side : the left part of the ship

6. Starboard Side : the right part of the ship

7. Inboard : the direction towards the center

8. Outboard : the direction away from the center

9. Hull : the plant building

10. Deck : the building floors

11. Main Deck : the cover of the hull

12. Superstructure : the structure above the main deck

The TMSR 500 overview is shown in Figure 2.3, Figure 2.4, and Figure 2.5.

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HP = High Pressure
LP = Low Pressure
SGC = Steam Generator Cell
SHXC = Secondary Heat Exchanger Cell

Figure 2.3 TMSR 500 Overview

TMSR 500 Plant and MSR Safety Features Figure 2.4 TMSR 500 Cutaway from Starboard Side

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The detail of each part of the ship is shown in Table 2.1.
Table 2.1 Part of the Shipyard Elaboration and Illustration

Part of the Sections Illustration
Ship
a. HP, IP and LP Figure 2.6 Plan View of Turbine Hall at TG Deck
Turbine Hall turbines Level
at TG Deck
Level b. Main generator Figure 2.7 Plan View of Turbine Hall from
c. LP and HP Starboard Side
Turbine Hall
Looking Feedwater Figure 2.8 Plan View of Turbine Hall from Port
Starboard heaters, deaerator Side
d. Sentry turbine and
Turbine Hall generator
Looking Port e. Auxiliary boiler

a. 5 decks in total
b. 4 HP feedwater

heaters on 4
decks
c. 3 LP feedwater
heaters on 3
decks
d. Deaerator on
main deck
e. Main turbines on
TG deck
f. Feedwater pumps
on low deck
a. Sentry turbine
b. Desalination plant
c. Sea water pumps
d. Turbine hall fan
room
e. Condenser
f. LP & HP
feedwater heaters
g. Deaerator &
heater
h. Auxiliary boiler
chimney
i. Operation house

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Turbine Hall a. 4 feedwater Figure 2.9 Plan View of Turbine Hall from Aft Side
Looking Aft heaters on 4
decks
Condenser
Looking b. Deaerator, turbine
Forward hall fan room,
chimney, and
crane on main
deck

c. Main turbines on
TG deck

d. Steam pipes
shown in color

e. Ballast cells on
both port and
starboard sides

a. 2 steam driven Figure 2.10 Plan View of Condenser from Forward
feedwater pumps, Side
3 hot well pumps,
desalination plant
and condenser on
low deck

b. Main generator,
exciter, sentry
turbine & generator
on TG deck

c. 21 to 26 kV bus
ducts and
generator circuit
breakers on TG
deck

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GIS Hall a. 3 main
Looking transformers and 1
Starboard spare transformer
on TG deck
Figure 2.11 Plan View of GIS Hall from Starboard
b. 420 kV HV circuit Side
breaker bay on TG
deck

c. 2 black start diesel
generators above
main deck

d. Diesel and water
tanks on low deck
(not shown)

Plan View of a. HV power lines go
GIS Hall out from the stern

b. All systems inside
double layer hull

Figure 2.12 Plan View of GIS Hall

HP = High Pressure
IP = Intermediate Pressure
LP = Low Pressure
TG = Turbogenerator
HV = High Voltage
GIS = Gas Insulated Switchgear

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The plant is divided into two parallel Power Modules (PMODs) in the forward
section. Each PMOD has the exact same layout. The Balance of Plant (BOP),
which is located behind the two PMODs, is the only remaining component of the
plant. There is some extra space of the plant to contain additional steam from
some of the systems of the nuclear island in case of leakage. Thus, it does not
need to be released into the atmosphere. Each plant cell is separated by thick
steel to ensure that if something happens in one of the cells, it does not affect
the other cells.
The power plant will be installed at the shipyard. It will be towed to the
deployment site and then sit onshore. As a result, everything can be
manufactured in a workshop and on-site construction can be minimized. There
is a lot of on-site reactor construction that takes a long time to complete. This will
increase the cost of power plant construction and cause a delay in the schedule.
The dimension of the plant is shown in Figure 2.13. The total height of the ship,
excluding the chimneys and cranes, is approximately 33 meters. The
circumference is approximately 67 meters. Thus, it is a large ship, but it is very
compact compared to a conventional nuclear power plant.

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GIS = Gas Insulated Switchgear
SHXC = Secondary Heat Exchanger Cell
SGC = Steam Generator Cell

Figure 2.13 TMSR 500 Sections

TMSR 500 Plant and MSR Safety Features

The energy flow diagram of the loops is depicted in detail in Figure 2.14. During
normal operation, the highest temperature is 704°C. The heat is transferred to
the secondary salt by the primary fuelsalt. The secondary salt is very similar to
the primary salt, but the ratio of the NaF and BeF2 is different and it does not
contain uranium, thorium, and all actinides. The secondary loop has slightly
higher pressure than the primary loop. The primary salt is unlikely to leak into the
secondary salt if there is a leak in the heat exchanger wall or tube. As a result,
the radioactive material will be retained in the primary loop, with fewer chances
of migrating to the secondary loop.
The TMSR 500 also has a third loop known as the tertiary loop. The tertiary loop
makes use of regular solar salt. It has been widely used in concentrated solar
power plants. The tertiary loop is not radioactive because it did not contain any
radioactive material and was not irradiated by any radioactive source.
There is almost no radioactivity on the steam-generator cell. As a result, the
steam is not radioactive, which means that nothing in the ballast plant is
radioactive. The steam temperature is approximately 550°C, which can be used
as supercritical steam. It is not extremely high because today’s most advanced
coal power plant can achieve steam temperatures of over 600°C, making all of
this feasible. All of these components are available from a manufacturer on the
market.

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2.2. TMSR 500’s Safety Philosophy

The safety philosophy of the TMSR 500 is based on the three Cs: Control, Cool,
and Contain. This concept is consistent with the IAEA's primary safety functions.
The IAEA defines the fundamental functions as reactivity control, heat removal,
and radioactive material confinement, which correspond to the three Cs used by
TMSR 500 respectively. The three Cs are critical, and there are several
approaches to achieving them. The system should implement active and passive
systems, an inherent safety system, and rely on human actions. The TMSR 500
relies on active, passive, and inherent safety systems. This will minimize human
error, which could lead to accidents in most industrial processes.
2.3. Safety Features of Molten Salt

During regular operation, the temperature is between 700°C and 800°C.
Because of the large margin between the operating temperature and the boiling
temperature of the molten salt, the system does not need to be pressurized. The
reactor can be operated at low pressure with no dispersal energy and no phase
change. The fuel salt chemically binds the harmful fission products. As a result,
the dangerous fission products are not released from the core even if there is a
leak. The large margin also allows the system to raise the heat up to 1000°C
safely.

The fuelsalt also has some beneficial chemical properties. There are no violent
reactions between TMSR 500 fuelsalt and air, water, or other substances. Even
if the molten salt is exposed to air and water, it will not catch fire because it does
not emit flammable or explosive gases. In some cases, high-temperature water
can violently react with zirconium to produce hydrogen, but there is no such
mechanism in molten fuelsalt. Even though molten salt is corrosive in some
situations, it does not react violently due to its slow reaction rate.

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Another benefit of using liquid fuel salt is that off-gas removal can be done online.
This system allows an online removal of unstable elements from the fuel salt
during regular operation, like the noble gases (xenon and krypton). As a result,
there was less accumulation of volatile fission products in the fuel. There are
fewer flammable fission products to release into the environment in the event of
an accident. In addition, the fluoride fuel salt has a high thermal capacity, and
the temperature rises more slowly than a coolant with a lower thermal capacity.
As a result, the possibility of a core meltdown, which most conventional nuclear
power plants fear, is eliminated in TMSR 500.
2.3.1 How to achieve control ?
The MSR 500 has three independent methods for controlling reactivity:

1. Total temperature coefficient of reactivity for TMSR 500 is negative.
Although not every element has a negative temperature coefficient of
reactivity, the sum of the values will always be negative.

2. The TMSR 500 is equipped with shutdown rods, which are a traditional
and highly reliable method of shutting down a reactor.

3. The TMSR 500 has freeze valves and a drain tank. The geometry of the
drain tank ensures that the fuelsalt cannot remain critical.

The TMSR 500 and a typical Gen III Light Water Reactor (LWR) control methods
are nearly identical. The difference is that if the shutdown rods fail, the LWR will
inject neutron poison into the coolant, whereas the TMSR 500 will drain the fuel
salt into the drain tanks.

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Figure 2.15 Freeze Valve and Drain Tank
Figure 2.15 depicts the TMSR 500's freeze valves and drain tank. As the
temperature rises, the freeze valve opens automatically. It requires no action,
and the fuelsalt flows into the drain tank by gravity without the use of any power.
The freeze valves and drain tank performance has been tested when the ORNL
operates their MSRE, but TMSR 500 is also intended to test the freeze valves
and drain tank on the pre-fission testbed platform.
2.3.2. How to achieve cool ?
The TMSR 500 has a normal residual heat removal powered by the sentry turbine
during normal operation. If either diesel or electric power is lost, the coldwall and
cooling pond can be used to keep the fuel cool for 145 days without using the
grid or diesel power. If all systems fail, the basement water and ballast tank can
keep the fuel cool for 269 days without using the grid or diesel power. The decay
heat will be low enough on the day 269.

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Safety System

TMSR 500 Plant and MSR Safety Features

The emergency core cooling system on the TMSR 500 is less complicated than
on the Gen III LWR, which also has a normal residual heat removal system
powered by either the grid or diesel. Because LWR is a high-pressure system, it
is difficult to inject water into the system once the coolant has been lost due to
the high pressure. Typically, LWR will have a complicated high-pressure,
intermediate-pressure, and low-pressure injection. LWR will most likely be
without power for three days, which means the emergency response must be
completed within those three days. However, TMSR 500 can be set to either 145
or 269 days, giving the operator plenty of time to take action carefully and
thoroughly.
The TMSR 500 has three methods to achieve the cool:
a. Sentry Turbine
Each TMSR 500 will be equipped with a 15 MWe sentry turbogenerator and a 50
MWth auxiliary boiler. The sentry turbine is always on during normal operation
and receives steam from the steam generator. The steam is bypassed and goes
to the sentry turbine rather than the main turbine. After the reactor is turned off,
the decay heat continues to generate steam in the steam generator, which can
keep the sentry turbine running for several hours. When the decay heat is low
enough, the steam will stop and the auxiliary boiler will be used to power the
sentry turbine.

Figure 2.16 The Sentry Turbine of TMSR 500

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Safety System

TMSR 500 Plant and MSR Safety Features

b. Coldwall and cooling pond
Figure 2.17 displays the TMSR 500's safety system known as the coldwall and
cooling pond. If there is no external power and the sentry turbine is not running
for some reason, this system will continue to cool the fuelsalt. Figure 2.16 shows
that the Can is inside the coldwall and that the coldwall is filled with water, so
heat from the Can and the fuel drain tank will radiate to the coldwall, and there
will also be some convection and conduction, but radiation will be the dominant
mode of heat transfer. The water in the coldwall is warmed and flows into a heat
exchanger that sits in the cooling pool; once cooled, it returns to the cold water,
and there is natural water circulation. The coldwall concept is not a new
technology; rather, it is a proven technology. It has been used in other advanced
reactors. For example, the HTGR uses coldwall to keep their vessels cool when
the power goes out. The coldwall concept is also used on the molten salt test
reactor by SINAP in China.

Figure 2.17 Coldwall and Cooling Pond
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Safety System