Introduction To ThorCon TMSR500 Technology and Passive Safety System

TMSR 500 Plant and MSR Safety Features

c. Basement Water and Ballast Tank
In case natural circulation does not form for whatever reason, the water in the
coldwall will be heated, but the TMSR 500 is equipped with a basement water
and ballast tank, as shown in Figure 2.18. Water in the silo hall basement will
cool the drain tanks and the cans, and this water can cool for 269 days. The
steam that generated by the heat will be quenched into the ballast tanks,
eventually.

Figure 2.18 Basement Water and Ballast Tank
2.3.3. How to Achieve Contain ?
The first barrier to a radioactive release is fuelsalt. The fluoride fuelsalt is the first
barrier for TMSR 500, and the oxide fuel matrix is the barrier for a typical LWR.
When compared to the oxide fuel matrix, fluoride fuelsalt performs significantly
better. This is because the off-gas system will continuously remove the noble
gases and some of the iodine that cannot be contained in the fuelsalt.

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Aside from the chemical features, there are also physical barriers shown in
Figure 2.19:

1. Stainless steel Can and the FDT that made of 25 to 50 mm steel;
2. Stainless steel Can Silo that consist of multiple 25 to 30 mm steel layers;

and
3. Stainless steel Can Silo Hall that consist of double 25 mm steel layers with

3 m concrete in the middle to defend against external impacts, such as
airplane crashes.

Figure 2.19 Multiple Steel Barriers

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SUMMARIES
1. The TMSR 500 is a modular reactor with several cores.
2. The TMSR 500 has four cores, but only two of them are active at any

given time, with the other two in cooldown mode.
3. Besides the primary and secondary loops, TMSR 500 also has a third

loop called the tertiary loop, which is separated from the fuel loop to
ensure that the generated steam is not radioactive.
4. TMSR 500 has 3 safety philosophies which are called the three Cs that
consist of Control, Cool, and Contain.
5. Three methods to achieve control:
a. Negative total temperature coefficient of reactivity
b. Shutdown rods
c. Freeze valve and drain tank
6. Three methods to achieve cool:
a. Sentry turbine
b. Cold wall and cooling pond
c. Basement water and ballast tank
7. Four barriers to achieving containment:
a. Fluoride fuel salt
b. Stainless steel Can and fuel drain tank
c. Stainless steel Can silo
d. Stainless steel Can silo hall

YOUTUBE VIDEO OF THE TRAINING RECORD:

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Chapter 3. Neutronic and Thermal-Hydraulic of TMSR
500 Plant

3.1. Current Status of Reactor Can Design of TMSR 500 Plant

The reactor Can design has undergone several changes between the previous
and current designs. The difference between the previous and current designs is
depicted in Figure 3.1. The previous design, shown on the left, has the heat
exchanger running vertically adjacent to the core, and the core is taller than it is
wide. At that point in the design cycle, the Oak Ridge National Laboratory (ORNL)
assumptions were used, but after conversing with graphite vendors, that
assumption was determined to be unfeasible unless a large sum of money was
spent.

Previously Currently

Figure 3.1. Comparison of the Old and Current Can Design of TMSR 500
[IAEA2020]

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The current design has roughly the same volume, but the Pot had to be shorter
and broader due to graphite constraints. As a result, the heat exchanger is
elevated above the reactor core, increasing natural circulation and making the
U-shaped heat exchanger more suitable for thermal expansion. However, it has
a disadvantage in that the salt volume is greater outside the core than inside.
3.2. Reactor Can Components
Figure 3.2 depicts the Can components. The Can consists of Pot, header tank
and primary pump, primary heat exchanger, uranium fuel makeup tank, thorium
fuel makeup tank, and freeze valves. The Can life is four years. It is constrained
by graphite degradation in the Pot.

Figure 3.2 Inside the Can
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Table 3.1 Characteristic of the Can

Output 557 MWth

Fuel NaF-BeF2-ZrF4-UF4-ThF4

Enrichment 19.75%

Moderator Graphite

Structure Stainless Steel

Outlet Temp. 704°C

Inlet Temp. 576°C

Fuel Boiling Temp. ~ 1400°C

Fuel Liquidus Temp. ~ 500°C

Operating Time 4 years

The characteristics of the Can are shown in Table 3.1. The fuelsalt composition
differs from the previous design of the TMSR 500. ZrF4 is included in the new
design. The addition of ZrF4 is intended to reduce corrosion and prevent uranium
oxide formation. The ZrF4 concentration in the salt is expected to be around 1%.
The addition of zirconium can improve the development of the TMSR 500
purification process by allowing many of the impurities in the fuel to precipitate
out.

The boiling temperature of fuelsalt is approximately 1400°C. It is an
approximation because the fuelsalt will change during the operation as the burn-
up changes. The liquidus temperature of the fuelsalt is around 500°C. There is a
big margin between the normal operating temperature and the boiling
temperature, no pressure is required to compensate for the high temperature.

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The Component of the Reactor Can include:
a. Reactor Pot
Figure 3.3 shows the reactor Pot and the fuel log in old design. This is not
the latest version of the fuel logs. The numbers might be different, but all the
phenomena and the general behavior apply to the present TMSR 500 reactor.
The old hexagonal log design is constituted by graphite as moderator in green
and fuel slot that shown as the small gaps in red. The fuel is heated through
along the three-or-four meters logs from bottom to top and reaches the maximum
temperature due to the fission.
The graphite is also heated by gamma and neutron irradiation so the fuelsalt is
also used to be the main coolant for the graphite. Figure 3.2 depicts the old Pot
reactor and the fuel log design. Although the numbers may differ, all phenomena
and general behavior apply to the current TMSR 500 design.

Figure 3.3. The old design of TMSR 500 Core and Fuel Log

Note: This configuration of moderator and fuelsalt is not used anymore in the
current design. But, the log shape would probably remain to be hexagonal.

The thermal expansion of the fuelsalt and the heating of the fuelsalt results in
two distinct feedbacks. Doppler broadening is related to resonance broadening
in fertile materials, specifically U-238 and Th-232. The broadening of the
resonance in the resonance region increases the probability of capturing while
significantly reducing the possibility of neutrons thermalizing and producing
fission. Fewer thermal expansions have different effects.
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The effective density of the fuelsalt inside the core will be reduced as the fuelsalt
is expanded. There are some variations of the reactivity feedback coefficient, but
it is always negative, around –2 pcm/K. This value also includes the graphite
reactivity feedback.
Thermal scattering and graphite thermal expansion are both affected by graphite
heating. The reactivity feedback coefficient is in the range of –2.7 pcm/K. The
first effect is related to neutron thermalization, which occurs when the moderator
is exposed to neutrons. It results in an effective equilibrium between the
thermalized neutron and the graphite temperature. Heat in graphite shifts the
Maxwellian distribution of thermal neutron equilibrium, effectively shifting the
neutron thermal spectrum. In addition, the graphite expansion effectively pushes
a small amount of fuelsalt out of the core. Both effects are added up and have a
negative overall effect in the current design. This is a reasonable number that
varies slightly depending on the specific core configuration.

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VIDEO 1. The Distribution of the Fission Power

The QR above will redirect to the website of Milano Multiphysics that shows a
Computational Fluid Dynamics (CFD) simulation of the fission power distribution.
The distribution is expected to follow a cosine towards the Z-axis and a back cell
function toward the radius in a common cellulose cylindrical reactor. The power
density is highest in the center, with lower power density in the bottom, top, and
radial periphery. The control rods were inserted in the center of this design, and
the power distribution was reduced.
The temperature distribution of the fuelsalt is calculated using the standard
power distribution. The power distribution differs from the temperature
distribution of the fuelsalt. In each log, the temperature distribution can be
adjusted for the flow rate. The temperature distribution is flattened as much as
possible to have a radially uniform outlet temperature and a uniform outlet
temperature of each log equal to the maximum fuel loop temperature.

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The graphite temperature is determined by the temperature of the salt which is
cooling it. The number of photons and neutrons that are used to heat it. As a
result, the graphite temperature is calculated by subtracting the local salt
temperature from the amount of energy released and dividing it by the effective
heat transfer coefficient between the graphite and the salt. The design of the
specific log is a trade-off between optimal moderation and optimal heterogeneity,
as well as the ability to sufficiently cool the graphite. Too much high graphite
temperature results in neutron damage that is faster than acceptable, which
consists of shrinking and then expanding again due to the defects and relocation
movement, but the graphite temperature can be analyzed similarly.

The profile shows significantly higher values in the upper part of the core, which
may not be advantageous for standard reactor design. There is no shift towards
the top of the reactor. Between the bottom and top, there is a fairly symmetrical
pattern. There is a slight shift toward the bottom because the cold salt on the
bottom side has a local higher reactivity in the sense that the higher density and
lower temperature result in a total higher local reactivity, whereas the slightly
higher average temperature on the top of the salt results in a slightly higher
Doppler Effect and negative reactivity. Because the electrical dimensional size is
not too large, the unbalance between the top and bottom is relatively mild.

b. Header Tank and Primary Pump
The purpose of the pump is to move the salt around the core. The circulation
time has changed in recent designs due to a more accurate estimation of the salt
capacity, but it is still in the range of 10 to 15 seconds. It means that the
movement around has to be more or less one cubic meter of salt per second,
which dictates the requirement of the pumps. The header tank is located at the
top of the primary loop and provides a surface area to remove off-gas from the
salt. It creates a space for salt at the gas-liquid interface.

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There is helium that is constantly being purged at a rate of 2 liters per minute.
One of the primary reasons for this is the ongoing elimination of Xe-135. Xe-135
is a massive neutron absorber that has a significant impact on uranium-fueled
thermal reactors. The Xe-135 is bound to remain within the fuel in a solid fuel
reactor. The behavior of these nuclides dominates reactor operation, particularly
when changing power. However, in the TMSR 500 reactor, this effect is
significantly reduced because the header tank can effectively remove the Xe-135
from the fuel. Depending on the specificity and porosity of the graphite, some of
it is absorbed. The effectiveness of the reactor operation in removing gas fission
products has a significant impact on reducing the source term in the event of an
accident.
c. Primary Heat Exchanger
The primary heat exchanger is made by a U-shaped shell and tube heat
exchange as seen in Figure 3.4. The adoption of the twisted tube is due to the
enhancement of the heat transfer coefficient on the shell side. The fuelsalt is on
the shell side and the secondary salt is on the tube side.

Figure 3.4. Primary Heat Exchanger of TMSR 500
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d. Freeze Valves
The freeze valve is a feature that has been used since the dawn of the MSR era.
The freeze valve is simply a pipe that is occupied by frozen solid fuel and serves
as a barrier to the flow of molten salt above it. The freeze valve operates
passively because it is continuously cooled by a helium flow that keeps the salt
plug cool. The heat in the capacity surrounding the valve begins to flow through
the valve to the plug of solidified salt until the salt melts and the valve opens as
soon as this active flow of helium is interrupted, either voluntarily or due to the
stopping of available power.
The design of the valves is directly from the Oak Ridge era. The operation of this
valve has already been tested in a real application and determined to be reliable.
The design also replicates well in terms of the time for the valves to open. The
precise time may not be known to the second, but it can be estimated and is
reliable. There are multiple valves in a total of four valves to ensure an acceptable
draining time for the transient. Four valves in the current design were modeled
after the Oak Ridge design, as shown in Figure 3.5.

Figure 3.5 Multiple Freeze Valve
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3.3. Neutronic and Thermal-Hydraulics Aspects
a. Drain Transient
The salt will be drained into the drain tank by gravity if the valves open due to the
absence of active helium cooling. The draining completes in approximately 10
minutes. The draining time depends on various factors, such as the amount, the
viscosity, and the density of the salt.

VIDEO 2. The Draining Transient of Molten Salt through Valve and Piping
System

The video above depicts a CFD simulation of a molten salt draining transient
through a valve and piping system. The actual shape of the core is not as shown.
The simulation employs a cylinder with the same volume of salt as the core, and
the other cylinder serves as the heat exchanger. The header tank is emptied in
the video above, and the pipe then completes the primary loop. Following that,
the primary heat exchanger is emptied, followed by the core. All of this happens
in about 600 to 700 seconds, and the salt flow is stopped after 700 seconds, and
all of the salt is in the drain tank, as shown in the video below.
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VIDEO 3. The Temperature Transient of Drain Tanks and Walls
The temperature transient of drain tanks and walls is shown in the video above.
The salt in the drain tanks primarily evacuates heat via radiative heat transfer.
Steel walls surround the drain tanks, which are effectively cooled and submerged
in water. Because the fuel factor from the drain tanks to the cool steel walls
exceeds 50%, the drain tanks release heat primarily through a relative heat
transfer and can grow rapidly based on the maximum temperature of the salt.
This refers to the previous design, but the difference with the current design is
minor and is only related to geometry, not behavior. Due to the convective
movement of the salt within the drain tank, the peak temperature is reached in
the salt in the drain tank just below the top free surface of the salt. The hot walls
radiate heat to the surfaces of the steel walls, which are kept cool by circulating
water. Water may reach a boiling temperature, and that nucleate boiling
accelerates heat removal.

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Figure 3.6 The Reactivity of the Core during the Draining
The reactivity of the core is reduced as a result of the draining. As shown in
Figure 3.6, when the draining line reaches the elevation level of the core, the
core begins to empty slowly but steadily. This will introduce negative reactivity,
resulting in the power being cut off. In most accidental scenarios, the reactivity
of the system is already severely negative during draining. The power is shut
down due to the introduction of negative reactivity, the control rod, and the
negative reactivity feedback caused by the salt heating up.

Figure 3.7 Drain Tank Creep Life Depletion in 1 Drain
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Figure 3.7 shows the decay heat in megawatts on the Y-axis and the time in
minutes after the core is shut down on the X-axis. Several minutes after the
shutdown, the power is in the megawatt range, and heat must be transferred to
the coldwall. This necessitates an increase in the temperature of the salt and
drain tanks. The creep damage of the drain tanks is now the primary sizing
criterion. The stress level is relatively low, but the temperature exhibits this
behavior as well, rising initially due to the higher decay heat and then slowing
and decreasing as the decay heat decreases. The creep lifetime is calculated
using the available creep correlation, which relates the maximum stress and
temperature. Creep is calculated in the positions with the highest stresses and
the highest temperatures. The creep damage in the drain tanks is relatively mild
in any relevant scenario; it is less than 1% of life.

k = 0.15642 k = 0.17549 k = 0.17550

Figure 3.8 Sketch View of Drain Tank.

The main concern with the drain tank is the possibility of its criticality in the event
of a leak that accidentally introduces water into the system. This is a likely
scenario, but the reactivity of the salt-filled drain tanks, even after mixing with the
makeup fuel, is far below one, so there is no risk of criticality even if the entire
space surrounding the drain tanks floods.

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b. Fuel make-up tanks
The fuel make-up tank holds the fuel required to feed the fission reaction
throughout the four-year operation. There is also a thorium make-up tank nearby
that to add extra thorium into the core. Because the fuel make-up tank is close
to the core, it receives a relatively mild neutron flux from the core that causes
fission in the make-up tank. In any foreseeable condition, the criticality of this
component is very low. Nonetheless, the heating caused by the fission in the fuel
make-up tank may cause a higher temperature than what is desirable for the
component. As a result, this component is shielded by steel, borated steel, and
boron carbide plates.

c. Effective delayed neutron fraction
The effective delayed neutron fraction change is one of the main effects that
distinguishes a circulating fuel reactor from a solid fuel reactor in neutronic
analysis. The effective delayed neutron fraction is the fraction of neutrons that
are not emitted immediately but are emitted with a certain delay during fission.
Uranium-235 absorbs approximately 0.7% of effective delayed neutrons. Some
of the delayed neutron precursors produced during fission are transported
through the core and beyond before decay when the fuel is circulating. Thus, a
large proportion of the delayed neutrons are emitted not in the pump, header
tank, or primary heat exchanger.

Nominal flow rates of the effective delayed neutron fraction are approximately
300 pcm. These numbers should not cause any concern because they are not
significantly lower than in the solid fuel fast reactor. The contribution of plutonium
isotopes and minor actinides in this case significantly reduces the delayed
neutron fraction. The peculiarity, however, is not the small value of the effective
delayed neutron fraction, but rather the fact that the delayed neutron fraction is
effectively dependent on the flow rate. In theory, the reactor can be run with a
variable speed pump and a variable circulation flow rate. When the pump is
turned off, the reactivity corresponding to the difference in β effective between
the static and circulating conditions is inserted. This is a well-known property of
molten salt reactors that has been studied since the Oak Ridge era. This is
ensured by TMSR 500, which takes it into account during analysis.

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d. Xe-135 reactivity
Xenon-135 is a powerful neutron absorber. When the reactor is not in normal
operation, the xenon level in a solid fuel reactor is determined solely by the
amount of I-135 produced as fission products of the decay chain. The amount of
Xe-135 depleted due to natural decay and neutron capture. The TMSR 500
reactor has a system that continuously removes gaseous fission product.
Calculating the effective removal of Xe-135 is difficult. The effectiveness of Xe-
135 removal through the surface gas to the interface must be defined and
confirmed. As well as the amount of Xe-135 that reaches the graphite pores. This
calculation can be hypothesized and assumed in two different scenarios:

a. The Xe-135 has a high rate of effective removal.
b. The Xe-135 has a low rate of effective removal.
After 15 days of operation, Xe-135 reaches an equilibrium value in both cases.

Figure 3.9 Effective Equivalent Xe-135 in Salt

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When the power is suddenly reduced from 100% to 50% on day 15, the
production of iodine drops from the equivalent equilibrium value of 100% to 50%,
and the fraction of iodine gradually decreases. Figure 3.8 depicts the effective
equivalent of Xe-135 in salt under both scenarios. If the xenon is removed at the
same effective rate, the system gradually adapts to the new level of xenon decay
rate caused by reduced iodine production. The initial slight increase is due to the
minor fraction of xenon atoms removed by neutron capture. The Xe-135 level is
almost constant at the start of the power change because it has not changed its
equilibrium value at all, and the fraction of xenon removed due to neutrons drops
by half. Figure 3.8 shows that the increase at the start is because the amount of
xenon removed by neutral capture in the TMSR 500 reactor is very small, almost
negligible.

Figure 3.10 Fast Effective Removal Rate
Figure 3.10 depicts the reactivity change caused by xenon during the 50%
transient (above) and the reactivity loss caused by xenon absorption (below).
The graph shows that the xenon effect is very small, less than pcm when
compared to the thousand pcm of PWR or HTGR. If the TMSR 500 system can
constantly tune the removal of xenon based on the time constant for the graphite
to migrate back to the salt, the TMSR 500 system can ensure a nearly zero
reactivity transient. The xenon can be handled regardless of the effective fraction
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of xenon that goes into graphite. The fraction of xenon that will permeate the
graphite is not precisely known from the value of the graph. The graphite is
treated to close as many pores as possible, but even if some xenon go into the
graphite, the effect will be much milder than in a solid fuel reactor. There will be
no forced shutdown period following an accidental change in power due to the
xenon peak, as there is in another reactor.

e. Off-gas Removal System
Fuelsalt is constantly circulating. When the fuel is transferred to the header tank,
an interface forms between the fuel liquid salt and helium that flowing through
the system two liters per minute. This allows the fission gases to move from the
salt to the gas. The required time for the gas fission product to leave the header
tank is quite long. As a result, a lot of decay occurs inside the header tank but
not in the core. Most of the known gaseous daughters of gaseous fission
products decay inside the header tank and return to the salt as soluble fission
products. Once the gas leaves the header tank, it goes to the Off Gas
Recuperators (OGR).

Each OGR is divided into four sections, resulting in a plug flow that prevents gas
mixing inside the OGR and increases adequate residence time. In this case, the
decay in released fission products is known as the gaseous daughter in the OGR,
which will deposit on the surface and efficiently radiate decay heat outside. Thus,
before returning to the system, the gas follows a long path to ensure that all
fission products have been captured.

The Can contains a header tank, OGR-1 and OGR-2. The radioactive gases are
noble gases when they leave OGR-2, and their daughters are not radioactive.
Therefore, if all of the radioactive daughters are contained within the Can, the
only thing left outside is noble gases, excellent dispersants with almost no
residence time in anything. Thus, the majority of the off-gas danger is contained
within the can.

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f. Decay Heat
In the event of an accidental release, a liquid-fueled reactor has the advantage
of being able to continuously remove the main source of radioactivity. This is
related to the removal of gaseous fission products, but it also implies that the
decay occurs outside of the fuel.

Figure 3.11 Decay Heat after 15 Days Operation at Nominal Power

According to Figure 3.11, the distribution of decay heat in the first few days after
shutdown is in the fuelsalt in the primary heat exchanger, where noble metals or
insoluble fission products may plate out. The header tank, first OGR, and other
gas components have very little decay. This is a conservative assumption from
Oak Ridge era analysis assumes that all unsolvable fission products plat out. Not
all of the noble metals were in the Primary Heat Exchanger (PHX). Nonetheless,
this necessitates consistent activation of the decay heat handling for this
component. The decay heat in the OGR is quite large, but it is well within the
manageable range of less than 1 MW. The TMSR 500 begins with less than 100
kW and quickly drops too much lower values.

Another advantage of the current TMSR 500 design is that the decay heat is
released immediately after an accident. The average energy of the emitted
gamma varies significantly with time after shutdown. Up to 40% of the decay heat
in the core is released into the graphite after shutdown. Therefore, within the 10
minutes between an accident and the adequate opening time of the freeze
valves, the graphite may absorb a significant portion of the energy of the decay
heat. Initially, the graphite has a higher temperature than the salt. However, due
to the faster decay, the salt temperature exceeds the temperature of the graphite.
These are extremely effective at lowering the maximum peak temperature of the
drain tanks after an accident.
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3.4. Principles of Fuel Management
The principles of fuel management for one power module (250 MWe) shown in
the Figure 3.12 below.

Figure 3.12 Principles of Fuel Management for One Power Module
The system began with Can A. There is a primary loop that is barely critical and
is filled with fuelsalt. The make-up salt is added at a rate of about 1 mL per
minute. Over time, this will add up to a total of 2 m3. There is no need to put
thousands of pcm of excess reactivity into the reactor at the start of the operation
with this make-up salt. The amount of make-up salt added can be adjusted based
on the amount required. It is capable of compensating for a wide range of
uncertainties in the core during operation.
The fuelsalt will be drained into the FDT after four years. For the next 2 – 4 weeks,
the fuelsalt will be in the drain tank. During that time, major maintenance work is
being carried out, particularly turbine maintenance. The fuelsalt is then
transferred from the FDT to Can B, where it will operate for another four years.
Can B also has a fuelsalt make-up tank that will supply additional make-up salt
as needed. When the salt has cooled sufficiently, it will be transferred to the
holding tank, which is similar to the drain tank. It is a cooling mechanism that
radiates heat to a water-filled coldwall. With a separate radiator circuit, the water
naturally circulates into the pond water. It can remain there for several months to
several years.
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The fuelsalt must be transferred from the holding tank to the vault before the next
draining event. The vault consists of six cylinders that look similar to the other
tanks but are larger and taller. It is 20 meters tall. The vault is referred to as the
TMSR 500 dry cask storage model because it is cooled by air rather than water.
This vault can hold spent fuel for up to 80 years. The entire movement of the
fuelsalt to the vault takes place inside double-walled pipes in the basement,
which are secured and safeguarded. The fuelsalt will be removed from the fault
and shipped to the final disposal once the permits are issued and the facilities
are built.
Based on previous experience in the United States, it is difficult to predict when
the final disposal will be ready. As a result, the reactor system must prepare
enough storage at the site that could last for the entire life of the power plant.
The TMSR 500 will include a re-processing system for removing uranium before
the fuelsalt is shipped to the final disposal location. The uranium in the used
fuelsalt is immensely useful because it is still significantly enriched, up to 9%.
The other goal of the re-processing is to extract beryllium, which is the next most
expensive component in the used fuelsalt. There will be any additional re-
processing but it does not depend on the technical issue but the political decision.

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SUMMARIES

1. Important components in the Reactor Can have been revised including
the reactor pot and primary heat exchanger.

2. The reactor pot uses graphite as a moderator and NaF-BeF2-ThF4-UF4 as
the fuel. The pressure is 3 bars with the inlet and outlet temperatures are
565°C and 704°C, respectively.

3. The lifetime of a reactor Can is 4 years because of the limitation of
graphite moderator in high neutron energy irradiation and high
temperature. The graphite will shrink then swells.

4. The reactor core has negative reactivity coefficients from fuel salt
temperature and graphite temperature.

5. The reactor will be shut down by rods, then the fuel is drained to the fuel
drain tank (FDT).

6. The purpose of the drain tank is to remove decay heat fully passively,
indefinitely.

7. A coupled neutronic-thermal-hydraulic/CFD model (Serpent and Open
FOAM) is implemented to fully describe the reactivity as a function of the
free surface height of the fuel salt.

8. The circulation time in the primary system is about 10 - 15 s.
9. Full draining takes about 786 s (13.1 minutes).
10. The subcritical condition is full filled for the make-up tanks and FDT when

those are flooded by water.
11. Freeze valve is a passive system.
12. FDT is cooled by the natural circulation of the cold wall.
13. Some drain transients have been calculated as well as the xenon behavior

due to the power operation or the effectiveness of off-gas cooling tanks.
14. The decay heat is deposited in the graphite inside the core, for example,

in the square block geometry, approximately 40% of the decay heat is
released in the graphite.
15. The shutdown margin of rods is about 2000 – 3000 pcm.

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YOUTUBE VIDEO OF THE TRAINING RECORD:

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Safety System

Operation and Balance of Plant of TMSR 500

Chapter 4. Operation and Balance of Plant of TMSR 500

4.1. Operation of TMSR 500
a. Normal Operation
The IAEA defines two basic categories: the Operational States and Accident
Conditions, as shown in Figure 4.1.

Figure 4.1 Review of Plant States
The TMSR 500 fission island is equivalent to the reactive building of a traditional
power plant. As shown in Figure 4.2, each fission island has two power modules,
and each module has 2 Cans. At any given time, one Can is in power operation
while the other Can is cooling down. There is a shared secondary coolant loop
for two Cans within each power module. Following the secondary coolant loop is
the tertiary salt loop, also known as the solar salt loop, because the salt is also
used in concentrated solar power. There is also a steam generator that produces
steam and feeds it to the turbine.

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Figure 4.2 Review of Fission Island
On the fission island, there is also a fuel shipping cask vault. The fuel shipping
cask will be dropped into the vault whenever fresh fuel is transported. On top of
that, there is a radtank, which is a shield and is also a part of the containment
boundary. Because of the thick shielding, there are floors above the radtank
where people can walk during power operation.
b. Normal Operating Modes
The TMSR 500 is a modular design; each plant typically has four Cans. It means
that the reactors in the power plants can operate in various modes at any given
time. For example, the reactor Can may be in power operation, cooling down,
and have recently been replaced at the same time. As a result, the operation
mode is defined as the Can mode rather than the plant mode.

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Figure 4.3 Normal Operating Modes
1. Mode 1 – Cold Pre-Startup

a. Remove hatches
b. Remove Radtank
c. Install the new Can, connect all pipes and cables including 2 secondary

loop lines
d. Install the Fuel Shipping Cask
e. Re-install Radtank
f. Re-install hatches
g. Purge and fill systems with helium to remove O2 and H2O
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2. Mode 2 – Hot Pre-Startup
a. Heat up all loops and salt tanks
b. Perform leak-rate tests
c. Flush the loops with molten flush salt to remove oxides
d. Close freeze valves
e. Insert shutdown rods
f. Transfer fuelsalts to the Can and verify the volume
g. Close containment system boundaries
h. Remove Fuel Shipping Cask

3. Mode 3 – Startup
a. Heat up all loops and salt tanks
b. Perform leak-rate tests
c. Flush the loops with molten flush salt to remove oxides
d. Close freeze valves
e. Insert shutdown rods
f. Transfer fuelsalts to the Can and verify the volume
g. Close containment system boundaries
h. Remove Fuel Shipping Cask

4. Mode 4 – Transition
• Increase power from 10% to 40%

5. Mode 5 – Power Operation
• Load following power operation between 40% and 100%

6. Mode 6 – Transition
• Decrease power from 40% to 100%

7. Mode 7 – Hot Standby
a. Heat up all loops and salt tanks
b. Perform leak-rate tests
c. Flush the loops with molten flush salt to remove oxides
d. Close freeze valves
e. Insert shutdown rods
f. Transfer fuelsalts to the Can and verify the volume
g. Close containment system boundaries
h. Remove Fuel Shipping Cask

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8. Mode 8 – Hot Shutdown
a. Insert shutdown rods
b. Decrease pump speed to maintain temperature
c. Use decay power in the first several hours for plant operation
d. Use grid or diesel power for longer time plant operation
e. Reactor can restart from hot shutdown

9. Mode 9 – Cold Shutdown
a. Stop the pumps
b. Stop off-gas system
c. Open freeze valves
d. Drain fuelsalt to Fuel Drain Tank
e. Maintain subcriticality by geometry, maintain cooling by Coldwall, contain
fuelsalt
f. No power required
g. Safe shutdown mode

10. Mode 10 – Post-Operation Cooldown
a. Transfer fuelsalt to the other Can in the same PMOD or to Used Fuel
Hold Tank
b. Fill the Can with helium and isolate it
c. Cools the Can using Coldwall for 4 years

11. Mode 11 – Post-Operation Removal
a. Disconnect the Can
b. Removal the Can from the plant

12. Transition from Cold Shutdown
a. Shutdown rods were inserted
b. Transfer fuelsalt from Fuel Drain Tank to the Can
c. Add thorium salt to compensate the increase of U-233 due to Pa-233
decay
d. If an unplanned fuel drain happens, reactor is prepared for re-startup in
this mode

13. Mode 13 – Cold Pre-Startup Non-Fresh Fuel
• Same as Mode 1 except that Fuel Transfer Cask is not used

14. Mode 14 – Hot Pre-Startup Non-Fresh Fuel
• Same as Mode 2 except that fuel is transferred from the other Can in the
same PMOD

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These modes are describing several scenarios:
• A new Can using fresh fuel goes through mode 1 – 11;
• A new Can using non-fresh fuel goes through mode 13 – 14, and then
mode 2 – 11.

In case that the plant does not need the power, the reactor goes into mode 7.
The hot standby mode can be last for a long time. Whenever the reactor needs
to be back in power, there are two scenarios:

• Restart from the shutdown (mode 8) goes through mode 3 – 11;
• Restart from the draining goes through mode 12, and then mode 3 – 11.

This number of modes is more than the LWR modes because TMSR 500 is a
moderate design. Modes 11 to 14 are for the fuelsalt transfer, and modes 9 and
10 are just for the replacement of the Cans. Table 4.1 contains a detailed
summary of these modes.

Table 4.1 Summary of the Detailed Modes

Fission Fuel

Mode Power Temperature Fuel Location

(°C)

Mode 1 Cold Pre-Startup N/A N/A N/A Shipping Cask or the other
Mode 2 FDT
Mode 3
Mode 4 Hot Pre-Startup ≤ 0.95 0% 550 – 700 Shipping Cask or the other
Mode 5 FDT, then Can
Startup ≥ 0.99 < 10% 550 – 704 Can
Mode 6
Normal Transition 576 – 704 Can
Mode 7 from Startup to ≥ 0.99 < 40% 576 – 704 Can
Mode 8 Power Operation
Mode 9
Power Operation ≥ 0.99 > 40%

Normal Transition

from Power ≥ 0.99 < 40% 576 – 704 Can
576 - 704 Can
Operation to Hot

Standby

Hot Standby ≥ 0.99 < 10%

Hot Shutdown < 0.99 0% 576 - 704 Can

Cold Shutdown ≤ 0.95 0% < 815 Can, then FDT

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Mode 10 Post-Operation ≤ 0.95 0% < 815 FDT, then UFHT or the
Cooldown N/A other Can
550 - 700 UFHT or the other Can
Mode 11 Post-Operation N/A N/A N/A
550 - 700 FDT, then Can
Removal
The other FDT
Transition from The other FDT then Can
Mode 12 Cold Shutdown to < 0.99 0%

Startup

Mode 13 Cold Pre-Startup N/A N/A

Non-Fresh Fuel

Hot Pre-Startup ≤ 0.95 0%
Mode 14
Non-Fresh Fuel

4.2. Load Following Power Operation

The load-following of a power plant is enchanting. Some electricity sources, such
as nuclear power, can run continuously for an extended time. Meanwhile, others,
such as wind and solar power, are unable to function consistently because they
are not always available (intermittent). As a result, the power consumption does
not remain constant. The power output of the grid should be adjusted, and load-
following is a method for the power plant to adjust and feed consumer demand.

TMSR 500 has several pumps: a primary loop, a secondary loop, a tertiary loop,
and a steam loop pump. There is also a sea feedwater pump that bypasses each
loop except the primary loop. The pumps of TMSR 500 can be seen in Figure
4.5. The pump speed and the bypasses can be adjusted to follow the load.

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Figure 4.4 Load Following System
The first step in the load following process is adjusting the feedwater, hot-well
pump speed, and bypass flow rate for the steam cycle. As a result, because the
temperature of the steam is nearly constant, the adjusted one is flow rate rather
than temperature. For the primary loop, there is an exception. The flow rate in
the primary loop is not changed; instead, the temperature of the fuelsalt changes.
Because of the negative temperature reactivity feedback coefficient, the fuelsalt
temperature will match the need for the secondary loop, and the fission power
will follow.

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Figure 4.5 The Steam Cycle Figure 4.6 The Tertiary Salt
Temperature Curve Temperature Curve

Figure 4.7 The Secondary Salt Figure 4.8 The Fuelsalt Temperature
Temperature Curve Curve

Figure 4.9 Pump Speed and Bypass Curve

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• Figure 4.5. From 0 to 100% power, the steam temperature is maintained to
be nearly constant throughout the process.

• Figure 4.6. The solar salt temperature must be kept nearly constant with a
slight variation to match the steam temperature.

• Figure 4.7. The secondary salt temperature rises as the power is increased,
but the margin remains constant.

• Figure 4.8. At zero power, there are no temperatures different from the
fuelsalt temperature, but as the power is increased, the temperature
difference grows. With this temperature change, the reactivity and the power
will change. The highest fuel temperature is in the 4th power mode.

• Figure 4.9. The primary loop pump speed is almost constant throughout the
power change.

The power will be changed by adjusting the pump speed and bypass. The typical
power change rate is 1% of full power per minute, and the maximum power
change rate is 5% per minute. For example, increasing the power from 40% to
100% will approximately take an hour. If the system exceeds 5% per minute, the
neutron gamma detector that monitors the efficiency of the neutron and gamma
counts will activate the shutdown rods, which will stop each interaction.

4.3. Balance of Plant (BOP)

The power plant is built to generate reliable power. Therefore, the concern is how
to defend against an accident and make the plant efficient and cost-effective.
Even if it is not a safety concern, the BOP is critical for power generation. The
BOP does not contain any radioactive material. The BOP is divided into several
sections, the most important of which are a turbine hall and a GIS hall.

Figure 4.11 depicts the typical steam cycle of the TMSR 500. The steam
generated by the steam generator, which drives the high-pressure, intermediate-
pressure, and low-pressure turbines. Before being pumped into the low-pressure
feedwater heaters, the steam is routed to the condenser. That is the primary
cycle, and there are always bypass and other cycles to improve efficiency.

Two features that important are the sentry turbine that can generate 15 MWe of
power and the bypass. If the main steam generator fails, this sentry turbine can
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be powered by an oil boiler in the plant and keep the power in the hot standby
mode for a short period. Steam can also bypass all of the turbines and travel
directly to the condenser, where it is cooled by seawater. When there is an
excess of power or steam, it can be dumped into the condenser, which has an
efficiency of up to 46.4%. This figure is much higher than the LWRs. This is
primarily due to the higher temperature of the steam.

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ThorCon Steam Turbine Model with 8 Feedwater Heaters - New Base May 2018 File Name:TC 49 New Base May 2018 2018-05-23 2050
STG Model: Doosan-Skoda Super-Critical

Operation and Balance of Plant of TMSR 500HP Turbine Bypass IP Feedwater for bypass operationIP & LP Turbine Bypass Condenser Water for bypass operation

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Safety System 0.06 241 MW

269.7C 547.0C 5.6 bara 5.6 bara
279.4C
2,876 kJ/kg 3,557 kJ/kg 279.4C 3,020 kJ/kg
276.0 kg/s
354.4 kg/s 354.4 kg/s 3,020 kJ/kg
168 MW
255 bara HP Turbine IP Turbine 290.9 kg/s LP Turbine
92.0%
547.0C 92.5% 94.6%

3,324 kJ/kg 185 MW 171 MW

255 bara 425.4 kg/s 14.9 kg/s 516 MW
547.0C DA/#4 FWH 35 MW
Sentry Turbine (ST)

3,324 kJ/kg

425.4 kg/s 20.6 bara Deaerator 0.083 bara
455.4C 0.907 dry
15.0 MW 42.2C

3,369 kJ/kg 90.0% 2,355 kJ/kg 30.7 kg/s
232.8 kg/s From ST & FWPT
15.7 kg/s Deaerator Storage Vessel Condenser 576 MW

15.3 MW 20.6 bara CW in 30.0 C 38.3C

455.4C

3,369 kJ/kg

Steam Condenser 18.0 kg/s 16.3 bara 5.6 bara 5.6 bara 2.5 bara 0.30 bara
193.1C 0.96 dry
Generator 370.8C 156.4C 136.5C
2,854 kJ/kg 2,520 kJ/kg
874 MW 71.5 bara 38.8 bara 20.6 bara 3,200 kJ/kg 660 kJ/kg 574 kJ/kg 16.0 kg/s 11.9 kg/s

347.3C 269.7C 455.4C Desup dT =.0C 14.8 kg/s 425.4 kg/s 410.5 kg/s #3 FWH
39 MW
3,004 kJ/kg 2,876 kJ/kg 3,369 kJ/kg 0.0 MW 96.6C Hotwell Pump

44.4 kg/s 26.5 kg/s 18.0 kg/s 165.9C 1.0 bara
107.93 C
702 kJ/kg
2,692 kJ/kg
103.7 kg/s 15.3 kg/s 42.2C 44.9C
#2 FWH 307 kg/s 43.2 kg/s
278 bara 39 MW 177 kJ/kg

288.0C #8 FWH #7 FWH #6 FWH #5 FWH 126.5C #1 FWH
31 MW
1,269 kJ/kg 85 MW 57 MW 56 MW 47 MW 532 kJ/kg Drains Cooler

425.4 kg/s 306.8 kg/s

288.0C 246.1C 216.2C 186.3C 66.7C

278.5 bara 42.7C
48.2C
Desup FWH dT = .0C 160.4C Main Feedwater Pump

Desup FWH MW = .0 MW 694 kJ/kg From IPT1 Exhaust

251.6C 221.7C 191.8C 165.9C 425.4 kg/s 15.0 kg/s 102.1C 72.2C

dT Consensate:Feedwater 5.5C 5.5C 5.5C 5.5C 5.5C 5.5C 5.5C

dT in FW Temp per Heater 41.9C 29.9C 29.9C 29.9C HP Feedwater 29.9C 29.9C 24.0C

90.0% Load Setting 100.0%
66.7 Reactor Thermal Output Capacity 1,117 MW
Note: ST = Sentry Turbine 33.9C dT from #3 14.7 MW
PI = Power Island 2.0 MW
4.0C dT (pump) Power Island Auxiliary Loads Fission Island Thermal Losses 1,115 MW

FI = Fission Island 0.0 kg/s Feedpump Power 14,378 kW Thermal Power to Power Island 539 MW
48.3%
CW = Circulating Water (condenser cooling water) Condenser Once Through CW Pumps 1,768 kW
531 MW
CW calcs currently based on fresh water Hotwell Pump Power 214 kW Cycle Output (inc. ST) 529 MW

Subtotal 3.1% 16,359 kW Gross Steam Cycle Efficiency 47.4%
517 MW
Plant Auxiliary Loads Generator Output (inc. ST)
PI Net power at LV side of Tx 46.4%
0.01%
Salt Pump Aux Load 8,800 kW PI Net Eff at LV side of Tx

Other Aux Loads 1,000 kW Power for Sale at HV Level

Modelling by Generation Solutions Limited for ThorCon US Inc. PI Aux Loads 1,981 kW Net Eff of Power for Sale
Est Transformer Losses 2,096 kW
Subtotal Aux Loads & Losses 13,877 kW Energy Balance Error
Printed 23-May-2018 20:50

Figure 4.10 Diagram of the Steam Cycle

Operation and Balance of Plant of TMSR 500

SUMMARIES

1. Since the TMSR 500 is a modular reactor, the operation mode is based
on the Can mode rather than the plant mode.

2. Can consist of the reactor core and primary system (heat exchanger and
pump) including uranium and thorium makeup tanks.

3. There are 14 operation modes for the normal condition, such as:
− M1 - M2: Pre-startup modes (cold and hot)
− M3: Startup mode
− M4: Transition of Increasing power
− M5: Power operation (load following between 40%-100%)
− M6: Transition of Decreasing power
− M7: Hot standby
− M8 – M9: Shutdown (hot and cold)
− M10: Post-operation cooldown
− M11: Post-operation removal
− M12: Transition from cold shutdown to startup
− M13: Cold Pre-Startup and Non-Fresh Fuel
− M14: Hot Pre-Startup and Non-Fresh Fuel

4. The TMSR 500 can be operated by the load-following power operation.
5. Maximum power change rate is 5% of full power per minute, compared

with the typical other NPP with the typical power change rate is 1% of full
power.

YOUTUBE VIDEO OF THE TRAINING RECORD:

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Safety System

Off-gas, Fuel Transfer, and Instrumentation and
Control Systems of TMSR 500

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

5.1. Fuel Transfer System
The fuel transfer system began with the Fuel Shipping Cask, as shown in Figure
5.1. Once the fresh fuel in the shipping cask is ready, it is liquefied and
transferred into the Can. The fuelsalt will enter the Pot via the header tank, where
it will be infused with makeup fuel and thorium fuel. There is a route from the
header tank to the FDT. Once the volume reaches a certain level, it will overflow
and go out of the header tank to the FDT. This happens during the normal
operation.

Figure 5.1 Simplified Fuel Transfer Diagram
There are some scenarios for the fuel transfer:
1. Normal operation for four years' lifetime.
Because the Can has a four-year lifetime, the fuelsalt Can must be replaced after
four years of operation. However, because the fuelsalt has an eight-year lifetime,
it is still usable for the next Can. Therefore, the fuelsalt is pumped to the next
Can within the same PMOD, which is right next to the current Can.
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2. Normal operation for eight years' lifetime.
The replacement of the Can began with melting the freeze valve to drain all of
the salt from the core to the FDT. The fuel is removed before the FDT is replaced
with the Can by pumping it to a used fuel holding tank for storage.

3. The fuelsalt was inadvertently drained
If the system accidentally drains the fuelsalt and it is necessary to return the fuel
to the same Pot to continue the operation, the fuelsalt can be pushed directly
back to the header tank.

A closer look at the Fuel Transfer Components is shown in Table 5.1 below.

Table 5.1 Fuel Transfer Components

Fuel Transfer In Can
Components

Figure 5.2 Plan View of the Can
Components

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Figure 5.3 Cross Section of the FDT.
Fuel Drain Tank

Figure 5.4 Plan View of the FDT
Components

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Out of Fuel Drain
Tank

Figure 5.5 Fuelsalt Lines Out of the FDT
When the fuelsalt is drained, a vent keeps the pressure constant. After stringing
the fuelsalt into the FDT, the fuelsalt is pushed out using pressurized helium
cover gas. The first step is to use the vacuum pump to transfer the fuelsalt from
FDT into the cruse. When the fuelsalt is in the cruse, the valve between the cruse
and the manifold opens, allowing the helium gas to press and push the fuelsalt
back to the manifold. The manifold also connects to several other components,
including the other Can, the used fuel holding tank, and the header tank.
The primary method of transporting the fuelsalt is by using helium gas supplied
by the off-gas system. However, because the helium gas has become attached
to the fuelsalt, it will be contaminated when the irradiated fuel is used. Therefore,
the gas must be collected and cleaned up by the off-gas system, and the gas will
then be recirculated back into the system.

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Off-gas, Fuel Transfer, and Instrumentation and
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5.2. Off-gas Systems

The off-gas system is unique to MSR. The off-gas system has a cover gas that
performs some important functions for the TMSR 500. The main functions of the
cover gas are:

1. Transfer the fuelsalt,
2. Dry seal of the pump, and
3. Remove the neutron poisons especially the xenon and krypton from the

fuel.

The fuel in a solid fuel reactor is usually sealed in the materials and cannot be
accessed. When the fuel burn-up increases, the fuel pins become pressurized,
and the fuel pin must be removed from the core at some point. If there are no
steel pins, the system can have a cover gas on top of the fuel, allowing the xenon,
krypton, and several other volatile fission products to be removed from the fuel
online. Helium was chosen as the cover gas for the TMSR 500 because it is
stable and cannot be activated, making it non-radioactive. Several main source
terms in the cover gas are noble gases like xenon and krypton, volatile elements
such as iodine, salt mist, and vapor, and some small particles which are fission
products that are not soluble in the fuelsalt. Once the cover gas is emitted from
the primary loop, it is also called the off-gas because it is contaminated.

The main functions of the off-gas system are:
1. Supply and circulate helium cover gas
2. Cool the off-gas and return most of the heat to the primary loop
3. Remove and contain radioactive isotopes
4. Separate krypton and xenon from helium
5. Remove other unwanted gases, mist, and particles

A process flow diagram is shown in Figure 5.6. The process starts with the
header tank and the pump that generates the off-gas. The thermal power shown
is the amount of heat generated by the off-gases in the header tank, which is 1.1
MW. Because the header tank is located within the Can, this heat will be returned
to the primary loop rather than being wasted. The dwell time in the header tank
is approximately 0.7 days. This is critical because the primary method of
removing the radioactive isotope is to allow it to decay. The longer the gas can
remain inside the component, the better.
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Off-gas, Fuel Transfer, and Instrumentation and
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Regarding that, the gas will be pushed to the Off Gas Recuperator no. 1 (OGR-
1) which has a much larger volume than the header tank. Because most of the
thermal power is generated by short-lived isotopes that have already decayed in
the header tank, the temperature drops and the thermal power decreases. The
off-gas will only be present in the OGR-1 for 2.2 days. The OGR-2 has the same
volume as the OGR-1, but the temperature and thermal power continue to fall as
more isotopes decay. The OGR-2 generates much less heat and has a slightly
longer dwelling time. The water will then pass through a filter, which will remove
residue particles and mist. When the temperature drops from 700°C to
approximately 200°C. The OGR-2 will absorb the majority of the fuel, including
salt mist and vapor. TMSR 500 have another filter that will remove residual
particles and mist.

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Off-gas, Fuel Transfer, and Instrumentation and Figure 5.6 Off-gas System Process
Control Systems of TMSR 500

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Off-gas, Fuel Transfer, and Instrumentation and
Control Systems of TMSR 500

In the SHXC cell basement, TMSR 500 will have two holdup tanks. The holdup
tanks are significantly larger than other tanks. They are approximately 20 times
the size of the recuperators. The off-gas will stay in this tank for a while to allow
the radioactive isotopes to decay. The majority of the short-lived ones will decay,
leaving the Can in the recuperators, while the longer-lived ones will decay out of
the Can in the holdup tanks, and then it will go to the charcoal bed, which can
retain the xenon for an extended time. The off-gas will enter the cold trap after
exiting the charcoal bed.

The cold trap aims to separate krypton and xenon from helium, which will be
returned to the system. Because krypton and xenon have much higher liquefying
and solidifying temperatures, they will separate from the helium and be pushed
to the collection bottles, while the helium will be pushed to the O2 and 3H Getters.
Its purpose is to remove oxygen as well as tritium from the system. The TMSR
500 utilizes nickel and titanium absorbers. After that, it will be transferred to the
recirculation compressor, and then back to the filter. This filter is capable of
removing particles larger than 1 micron in size. After that, it is pushed back to the
pump, through the dry gas seal, and back to the header tank via the flow
regulator, which controls the flow of the entire system.

All of these calculations, including the dwelling time and thermal power, are
based on a 2 liters per minute operating flow rate. The flow rate can be increased
or decreased as needed, but it will be kept at 2 liters per minute in normal
operation, which is sufficient for the operation of the system. There is also a
helium makeup bottle that provides fresh helium if more helium is required.
Figure 5.7 depicts a summary of the flow by functions in the normal operation
mode.

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Off-gas, Fuel Transfer, and Instrumentation and
Control Systems of TMSR 500

Figure 5.7 Off-gas System Process by Functions in Normal Operation
The TMSR 500 also has two other modes beside the normal operation mode:
1. Regeneration mode
The xenon and krypton are pushed out of the system and into storage bottles,
while fresh helium is constantly supplied to the system. The reactor does not
need to be shut down while in regeneration mode because the system can be
done online.
2. Purge mode
While fresh helium is being supplied to the system, the circulation of the system
can be flushed. The purging mode is intended to clean up the system.
The summarized of TMSR 500 for each mode flow is shown in Figure 5.8.

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Off-gas, Fuel Transfer, and Instrumentation and Figure 5.8 Flow Diagram of Each Modes
Control Systems of TMSR 500

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Off-gas, Fuel Transfer, and Instrumentation and
Control Systems of TMSR 500

The main components of the off-gas system are:
a. The Off Gas Recuperators
The Off Gas Recuperators (OGR) are the main components of the off-gas
system. Figure 5.9 depicts the design and features of the OGR. The purpose of
the design is to keep the gas in the recuperator for as long as possible. The flow
rate determines how long the gas can stay in the recuperator. There is a flow
regulator that controls the flow rate. As a result, the amount of the isotope that
will decay in the recuperators can be predicted. Inside the recuperators, there is
a material that will absorb the fission products, particularly the metals.
When the temperature drops, the mist and vapor condense in the recuperator.
The off-gas temperature is above 700°C when it first exits the fuel salt, but it
drops to around 200°C after passing through two recuperators. Therefore, the
majority of the mist vapor will be deposited on the surface of the recuperators.
Both recuperators are contained within the Can, so they were cooled by the
coldwall and replaced with the Can every four years.

Figure 5.9 Off-gas Recuperators
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Off-gas, Fuel Transfer, and Instrumentation and
Control Systems of TMSR 500

The TMSR 500 has an off-gas holding tank outside the Can that is very similar
in design to this OGR. The only difference is that the holding tanks are much
larger than the recuperators, and they are located in the basement of the SHXC
cell, which is flooded and cooled by water.
b. The Cold Traps
In the cold traps, the xenon and krypton are separated from the helium using the
temperature difference. Once the off-gas passes through the cold traps, the
xenon and krypton are solidified while the helium gas passes through the cold
trap. The accumulated xenon and krypton in the cold trap will be drained to the
evaporator when the reactor is in purge mode. After the draining is complete, the
V2 and V4 are turned off. As a result, the temperature will rise faster. After the
xenon and krypton are moved to the evaporator, the evaporator will be warmed
up to room temperature and the pressure will increase. In this case, the pressure
will be increased to approximately 40 bars. Figure 5.10 depicts the illustration of
the cold traps.

Figure 5.10 The Cold Traps

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