Molten Salt Reactor Experiment Research Articles

ZERO POWER CRITICALITY BENCHMARK EVALUATION OF THE MSRE IN WIMS

This paper discusses work done to benchmark the deterministic code WIMS [1] against the Monte Carlo code Serpent [2] and experiment. Comparison is made against the Molten Salt Reactor Experiment at Oak Ridge National Laboratory as well as a Serpent model produced at the University of California, Berkeley. Producing a model for an MSR is possible thanks to the development of the next version of WIMS, WIMS11. The structure of the WIMS model built is discussed, and the final predicted criticality value for the MSR is given. This compares favourably with the Serpent model; however, both codes predict values considerably different to those expected. Potential reasons for this are suggested. However, it is concluded that WIMS has successfully been benchmarked against the current state of the art. This provides confirmation that this is a valid approach for molten salt reactor research analysis.

Development of a steady state analysis code for molten salt reactor based on nodal expansion method

The unique feature of molten salt reactor (MSR) is that heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture which acts as both fuel and coolant. The fission energy from fissile fuel is released immediately into the fuel salt in the core, and the delayed neutron precursors (DNP) have a loss due to fuel circulation in the primary loop, which both make MSR be characterized by the strong interplay between the neutronics and thermal hydraulics (TH) and imply that the software developed for traditional solid-fueled reactors is not applicable to MSR. In this study, a new coupled neutronics and TH code named TMSR3D is developed for evaluating the steady-state characteristics of liquid-fueled MSR. In this code, the fourth order standard nodal expansion method (NEM) is implemented to solve the neutron diffusion equations, and the parallel multi-channel model is adopted to take into account the feedbacks of local temperature to group constants. Besides, a convection term is introduced into the traditional balance equations for DNP to consider the influence of fuel salt flow. Some experiment data of Molten Salt Reactor Experiment (MSRE) are chosen to verify the code, and the core characteristics including the distributions of neutron fluxes, DNP and temperature under the rated condition are simulated and discussed. Furthermore, the impacts of the inlet mass flow rate, inlet fuel temperature and the fuel salt residence time out of the core are analyzed in detail. The numerical results indicate that the TMSR3D code can provide a reliable description of the neutronics-TH coupling phenomena and of the steady state analysis for graphite-moderated channel-type MSRs.

A systematic approach to identify initiating events and its relationship to Probabilistic Risk Assessment: Demonstrated on the Molten Salt Reactor Experiment

One of the first steps in developing a risk assessment model is an exhaustive search for initiating events, which is a systematic and comprehensive starting point to answer the question “what can go wrong?” for a given system design. Identifying Postulated Initiating Events (PIEs) for a reactor design that is at a conceptual or preliminary stage facilitates the incorporation of risk insights into the next iteration of the design process and allows for the early establishment of more quantifiable risk assessment models, such as event sequence diagrams and event tree analysis. Liquid-Fueled Molten Salt Reactors (LF-MSRs) are an example of an advanced reactor technology that does not benefit from having a wealth of operating experience or prior risk-informed safety assessment efforts. Furthermore, design details, such as normal operating conditions and the composition of radioactive material inventories, can deviate substantially from those in other reactors, such that a systematic and comprehensive approach to identifying PIEs for an LF-MSR may highlight accident initiators that have not previously been identified. In the present work, the Master Logic Diagram (MLD) and Hazards and Operability (HAZOP) study approaches were used, together, to identify and consider PIEs for multiple inventories of radioactive material across various Plant Operating States (POSs) in a specific LF-MSR design -- the Molten Salt Reactor Experiment (MSRE). Potentially risk-significant PIEs identified during the analyses of the MSRE design are presented. Furthermore, considerations for exhaustively identifying PIEs for advanced reactor designs are discussed; for example, the combination of inductive and deductive methods was found to provide a robust identification of PIEs in a way that is conducive to the analysis of a nuclear reactor design at an early design stage.

A unique molten salt reactor feature – The freeze valve system: Design, operating experience, and reliability

Reliable mechanical valves that can withstand the corrosive and high-temperature conditions in Molten Salt Reactors (MSRs) have not yet been demonstrated. In their place, freeze valves (sometimes called freeze plugs) represent a unique nuclear design solution for isolating salt flow during operations. Additionally, in some cases, they are intended to perform safety-related functions. As implied by the name, operation of a freeze valve depends upon the controlled phase change of salt, and the ability of increased reactor heat from transients to induce modulation of the “valve” has frequently been equated to safety function passivity. In this article, a review of literature pertaining to freeze valve design, performance, and thermodynamic analyses is presented first. The Molten Salt Reactor Experiment (MSRE) was the first MSR to implement the use of freeze valves and is the only liquid-fueled MSR with significant operating experience; yet, the success of the MSRE freeze valve to thaw on demand or remain frozen during power operation depended on two subsystems comprising power-operated sensors, mechanical valves, and other active components. Accordingly, as executed at the MSRE, the freeze valve system was not a fully passive safety feature. Performance of original Process Hazards Analysis (PHA) studies on an MSRE safety-related freeze valve design indicates failure of multiple individual active components could result in the impairment of the freeze valve function, such that the freeze valve itself is a system of components – rather than a single component. Fault Tree Analysis (FTA) based on PHA results provides estimated failure rates for two important failure modes of freeze valves. The results suggest that designers of freeze valve systems may face a trade-off between the reliability of the freeze valve to thaw on demand and the reliability of the freeze valve to remain frozen during normal reactor operations; however, a sensitivity study showed that the likelihood of a spurious thawing of the MSRE freeze valve system could have been reduced without significantly affecting the likelihood of a failure to thaw.

A Review of Circulating Voids and Gaseous Fission Product Behavior in Molten Salt Reactors

Abstract Molten salt reactors (MSRs) are a family of advanced, generation-IV, nuclear reactors that use a circulating molten alkali fluoride fuel salt as both a primary working fluid and a fuel matrix. Since the fuel salt is liquid, gases become entrained in the fuel salt flow, and these entrained gases are called circulating voids. This paper reviews some aspects of circulating voids on reactor behavior. A formal definition of the circulating void fraction is given. A time-line of Oak Ridge reports on circulating void behavior is presented. Methods to measure the circulating void fraction are described. Foam induced by gas entrained in the fuel salt is discussed. The discovery of circulating voids in the molten salt reactor experiment (MSRE) is detailed. Aspects related to bubble size and interfacial area are reviewed. The Laplace pressure is examined. The void and pressure coefficients of reactivity are examined. Some requirements for a bubble model for xenon behavior analysis are proposed. The effects of bubble surface mobility are examined. Finally, sources of gas production in the fuel salt are investigated.

A dynamic model of xenon behavior in the Molten Salt Reactor Experiment

A dynamic model of xenon behavior in the Molten Salt Reactor Experiment (MSRE) has been developed. This model is the first to describe the behavior of both the start up and the shut down transients in the Molten Salt Reactor Experiment (MSRE) with a single parameter set.

Code development and analysis of heat pipe cooled passive residual heat removal system of Molten salt reactor

The high efficient and reliable Passive Residual Heat Removal System (PRHRS) is of great importance to the development of Molten Salt Reactor (MSR), whether the liquid fuel MSR or solid fuel MSR. In this research, based on the structure and operation performance of the new conceptual PRHRS, using high temperature heat pipe as heat transportation, a set of reasonable mathematical and physical models have been built, mainly including reactor core model, high temperature heat pipe model and PRHRS model. The new conceptual Passive Residual heat removal system Analysis Code (PRAC) has been developed using FORTRAN language. The verification and validation for PRAC is conducted by using existed Molten Salt Reactor Experiment (MSRE) data provided by Oak Ridge National Laboratory and the PRHRS experimental data in our previous work. The results show that the developed PRAC could predict the MSR transients reasonably and show high fidelity. The numerical results of reactor core neutronic and thermal–hydraulic model show a good agreement when compared with MSRE pump startup and shutdown transients. The numerical results by PRHRS model agree well with experimental data which the maximum error of the fluoride salt temperature and heat pipe wall temperature are within 1.2% and 2.8%. This research provides a valuable software basis for the design and analysis of PRHRS of MSR.

Sensitivity/uncertainty comparison and similarity analysis between TMSR-LF1 and MSR models

TMSR-LF1 (Thorium-based Molten Salt Reactor-Liquid Fuel) is the reactor chosen to be developed in the “Thorium molten salt reactor nuclear energy system” project of the Strategic Priority Program launched by the Chinese Academy of Sciences. The software and data library adopted in its design needs to be verified and validated by using benchmarks. In order to guarantee the validity of the verification results, the being-designed reactor should have similarity (index Ck and E ≥ 0.8) with the benchmark reactor. Using TSUNAMI module and in SCALE 6.1, the Sensitivity/Uncertainty (S/U) analyses and similarity assessments were performed in TMSR-LF1 and the candidate benchmark reactors: MSRE (Molten Salt Reactor Experiment), MSBR (Molten Salt Breeder Reactor) and the Critical Experiment Device (hereafter the device). Uncertainty was calculated using both SCALE and TMSR 44-group covariance data. It is found that the total eigenvalue uncertainties deduced from covariance data in TMSR-LF1 and the candidate benchmark reactors satisfy the ±1% design accuracy requirement proposed by M. Ishikawa in 2008, except in MSBR. To meet the accuracy of ±0.3%, targeted experiments for TMSR are needed. The covariance data of 7Li, C-graphite, 58Ni and 19F were the main contributors to the uncertainty of effective multiplication factor as well as the covariance data of fuel nuclides. The calculated similarity indices indicated that there was a good similarity (Ck and E > 0.9) between TMSR-LF1 and MSRE, and a lesser similarity between TMSR-LF1 and the device (Ck and E > 0.8). The uncertainties should be improved and more benchmarks for the direct verification and validation should be provided by relevant experiments.

Development and Verification of a System Dynamic Model for Liquid-Fueled Molten Salt Reactor Based on Modelica

The liquid-fueled molten salt reactor uses liquid fuel, and its dynamic characteristics are different from the traditional solid fuel reactor. Dynamic modeling and simulation play a key role in the design phase of the innovative reactor system. Based on Modelica language, a Modelica model named TMSRLib (Thorium-based Molten Salt Reactor Library) for dynamic modeling and simulation of liquid-fueled molten salt reactor system is developed using Dymola platform. In order to verify the validity of the developed dynamic simulation model library, two system models of the MSRE (molten salt reactor experiment) with core represented by single region and nine regions are built based on the TMSRLib Modelica library, respectively. Then the two system models of the MSRE are validated by using experimental benchmarks of MSRE. The results show that the numerical results are in good agreement with MSRE experimental results, and verifies the applicability and correctness of each component model in TMSRLib.

Plant-level dynamic modeling of a commercial-scale molten salt reactor system

Ongoing research by the authors has led to a dynamic modeling approach that was verified against experimental data from the Molten-Salt Reactor Experiment (MSRE). These lumped-parameter models, characterizing the reactor dynamics, are nonlinear and represent changes in mass, energy, and temperature in all parts of the reactor plant. The reactivity feedback effects due to changes in temperature and Xe-135 concentration are taken into account. These models can be used to study both the time and frequency response to perturbations caused during normal operating conditions and during anomalies resulting from failure of certain sub-systems. A plant-level dynamic model of a representative molten salt reactor system developed based on the same methodology is presented here. A once-through steam generator providing superheated steam and reheated-regenerative Rankine cycle balance-of-plant system are coupled to the plant model. Results from simulation are presented for various transients and the resulting response of the plant is analyzed. Frequency response of the plant is also presented for various boundary conditions on the secondary side. Modeling results suggest that the inherent temperature-related feedbacks result in load-following characteristics that can be leveraged to engineer control systems offering a great deal of autonomy.

Effect of FLiBe thermal neutron scattering on reactivity of molten salt reactor

Thermal neutron scattering data has an important influence on the calculation and design of reactor with a thermal spectrum. However, as the only liquid fuel in the Gen-IV reactor candidates, the research on the thermal neutron scattering effect of coolant and somewhat moderator FLiBe has not been carried out sufficiently either experimentally or theoretically. The effect of FLiBe thermal neutron scattering on reactivity of TMSR-LF (thorium molten salt reactor - liquid fuel), TMSR-SF (thorium molten salt reactor - solid fuel) and MSRE (molten salt reactor experiment) were investigated and compared. Results show that the effect of FLiBe thermal neutron scattering on reactivity depends to some extent on the fuel-graphite volume ratio of core. Calculations indicate that FLiBe thermal neutron scattering of MSRE (with the hardest spectrum) has the minimum effect of 41 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-SF (with the softest spectrum) has the maximum effect of -94 pcm on reactivity, and FLiBe thermal neutron scattering of TMSR-LF has an effect of -61 pcm on reactivity at 900 K.

Modeling and simulation functional needs for molten salt reactor licensing

The modeling and simulation functional needs to license and design molten salt reactor concepts were assessed and an evaluation methodology was developed using a licensing-focused approach considering the modeling and simulation components that contribute to a molten salt reactor safety analysis report. These functional needs were prioritized by identifying the relationships between functional needs and using a quantitative weighting method to identify the most critical components to build a licensing case. This novel weighting method provided a list of functional needs prioritized relative to key advanced reactor licensing outcomes. In addition, functional need inputs were ranked to reveal those that feed the most modeling and simulation components. The findings from this prioritization effort were used to inform the modeling and simulation goals of the US Department of Energy Office of Nuclear Energy (DOE-NE) Molten Salt Reactor campaign.Modeling and simulation needs were grouped into three main categories: (1)source term definition, material accountancy, and radiological consequences, (2)safety analysis, and (3)steady-state neutronics and thermal hydraulics. Five near-term goals were identified among these focus areas that add value: (1) deploying a systems and safeguards analysis computational tool, (2) defining source terms for molten salt reactor analysis and interfacing with the US Nuclear Regulatory Commission on accident progression analysis, (3) developing point designs for liquid-fueled thermal-spectrum and fast-spectrum molten salt reactors, (4) building safety analysis specifications and reference solutions for Molten Salt Reactor Experiment transients, and (5)interfacing with the thermophysical and thermodynamic salt chemistry databases.

CFD Simulations of Molten Salt Reactor Experiment Core

At present, no clear guidelines exist for modeling non-water-cooled small modular reactors (SMRs) despite the rising need for high-fidelity simulation tools to support regulators and the industry. Most SMR concepts currently under the Canadian prelicensing review adopted non-water-cooled–reactor technologies [molten salt reactor (MSR), gas-cooled reactor, and liquid metal–cooled reactor] that are new for Canada. There is a need for a modeling tool set that is broadly applicable for the assessment of advanced technologies used in SMRs. Computational fluid dynamics (CFD) can be used in performance evaluation and safety analysis of non-water-cooled SMRs for modeling three-dimensional (3-D) fluid flow and heat transfer in geometries of arbitrary complexity without resorting to geometry-specific empirical correlations. This study investigates the capabilities of existing models within a commercial CFD code to simulate the flow and heat transfer characteristics in a MSR configuration. The Oak Ridge National Laboratory (ORNL) Molten Salt Reactor Experiment (MSRE) configuration was simulated in this study using a stand-alone CFD approach, and CFD predictions were assessed with ORNL data. Intricate geometry details within the MSRE core were included in the computational model to study the associated geometric effects. The results obtained in this study showcased the ability of CFD to predict 3-D effects within the computational domain especially at the lower plenums. The predicted trends for the temperature rise in the fuel and moderator within the core were in good agreement with the ORNL data. The results presented in this paper constitute the first step in developing Canadian Nuclear Laboratories’ capability for CFD modeling of non-water SMRs.

Application of Monte Carlo method to calculate the effective delayed neutron fraction in molten salt reactor

Delayed neutron loss is an important parameter in the safety analysis of molten salt reactors. In this study, to obtain the effective delayed neutron fraction under flow condition, a delayed neutron precursor transport was implemented in the Monte Carlo code MCNP. The molten-salt reactor experiment (MSRE) model was used to analyze the reliability of this method. The obtained flow losses of reactivity for 235U and 233U fuels in the MSRE are 223 pcm and 100.8 pcm, respectively, which are in good agreement with the experimental values (212 pcm and 100.5 pcm, respectively). Then, six groups of effective delayed neutron fractions in a small molten salt reactor were calculated under different mass flow rates. The flow loss of reactivity at full power operation is approximately 105.6 pcm, which is significantly lower than that of the MSRE due to the longer residence time inside the active core. The sensitivity of the reactivity loss to other factors, such as the residence time inside or outside the core and flow distribution, was evaluated as well. As a conclusion, the sensitivity of the reactivity loss to the residence time inside the core is greater than to other parameters.

Development of a dynamics model for graphite-moderated channel-type molten salt reactor

A molten salt reactor (MSR) is one of the six advanced reactor concepts selected by the generation IV international forum because of its advantages of inherent safety, and the promising capabilities of Th-U breeding and transuranics transmutation. A dynamics model for the channel-type MSR is developed in this work based on a three-dimensional thermal–hydraulic model (3DTH) and a point reactor model. The 3DTH couples a three-dimensional heat conduction model and a one-dimensional single-phase flow model that can accurately consider the heat conduction between different assemblies. The 3DTH is validated by the RELAP5 code in terms of the temperature and mass flow distribution calculation. A point reactor model considering the drift of delayed neutron precursors is adopted in the dynamics model. To verify the dynamics model, three experiments from the molten salt reactor experiment are simulated. The agreement of the experimental data and simulation results was excellent. With the aid of this model, the unprotected step reactivity addition and unprotected loss of flow of the 2 MWt experimental MSR are modeled, and the reactor power and temperature evolution are analyzed.

A simultaneous corrosion/irradiation facility for testing molten salt-facing materials

Aside from the historical Molten Salt Reactor Experiment, a few in-reactor loops, and one electron irradiation/corrosion facility, dedicated facilities to test the combined effects of molten salt corrosion and irradiation on materials do not currently exist. A major gap therefore exists in rapid, reactor-relevant materials testing capabilities which, if remedied, would greatly hasten molten salt reactor development. We present a new accelerator-based facility for rapid, simultaneous testing of molten salt-facing materials utilizing a proton beam as the radiation source. Introducing proton irradiation to a molten salt corrosion system poses specific engineering concerns in sample and corrosion cell design, operational stability, integration with the accelerator beamline, and radiation safety. This paper describes how these requirements were fulfilled with confirmatory tests and results.

Thermal hydraulic model of the molten salt reactor experiment with the NEAMS system analysis module

System analysis codes have a long history of providing best-estimate and conservative safety analysis for both light water and advanced reactor technologies, including molten salt reactors. As interest continues to expand with advanced reactor concepts, system analysis codes will need revisions to accommodate the behavior of these technologies. Legacy system analysis codes will need to be updated to the latest numerical techniques to shorten execution time and increase the accuracy of results. One example of a modern system analysis code that already encompasses these characteristics is the System Analysis Module (SAM).One key objective of this paper was to review available information for system code modeling of the Molten Salt Reactor Experiment (MSRE) from sources in the open literature and collect the information from these open sources in one place for the first time. This supports the potential objective of developing an open specification for system code analysis for MSRE steady state and transients with and without reactor kinetics. Data from actual MSRE tests will serve as the basis for code-to-code comparison exercises, including the MSRE zero power physics tests, the fuel pump start-up and coast down tests, and the natural circulation transient. The objective is to produce a code-to-code benchmark with a standardized set of comparison problems, recognizing the limitations of the original data.To demonstrate an initial application of this objective and the usefulness of compiling this open data, two Molten Salt Reactor Experiment (MSRE)-related models were developed to evaluate SAM for liquid fueled molten salt reactors. One model was the SAM MSRE hydraulic mockup, which provided experimental data for pressure drop measurements. The second model was the complete MSRE primary loop. The MSRE primary loop model incorporated a fluoride salt fuel/coolant with heat transfer in both the core and heat exchanger. For both the hydraulic mockup and MSRE primary loop models, a holistic 1-D system description was built using open documentation, an open description that can be readily modified and applied for any system analysis code.SAM results for the pressure drop of the hydraulic mockup model were within 6% with measurements. Coolant temperatures for the primary loop model matched the expected axial change in temperature from historical calculations. Using alternative coolant properties obtained from the literature, corresponding to salts with different actinide contents, returned similar trends in core temperature profiles. A thermal hydraulic demonstration of a loss-of-flow transient showed the importance of coupling SAM thermal hydraulic analysis to neutronics. This coupling is essential for simulating MSR transients with system analysis codes.

Sensitivity/uncertainty analyses comparing LR-0 reactor experiments containing FLiBe salt with models for molten-salt-cooled and molten-salt-fueled reactors

Critical experiments using an insertion zone with FLiBe salt performed at the LR-0 reactor at Research Centre Řež (RC Řež) have been compared to application models for solid-fueled, fluoride-salt-cooled high-temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts using sensitivity and uncertainty (S/U) analysis techniques. These experiments support FHR and MSR advanced reactor concepts by informing on neutron spectral effects and nuclear data uncertainties related to fluoride salts. The FLiBe salt in the LR-0 experiments is from the Oak Ridge National Laboratory Molten Salt Reactor Experiment and is enriched to greater than 99.99% 7Li. This work is part of a broader collaboration between the United States and the Czech Republic on MSR and FHR technology. Results from the S/U analyses for FHR and MSR models indicated the most significant potential source of eigenvalue bias due to nuclear data within the FLiBe salt is radiative capture in 7Li. Other smaller but potentially significant contributions come from 19F, 6Li, and other 7Li reactions. Similarity comparisons of the RC Řež LR-0 experiments and FHR and MSR application models indicate that the LR-0 FLiBe experiments could be useful as candidate benchmarks for low-enriched uranium–fueled application models. However, sensitivity differences observed included both spectral effects, due to the LR-0 reactor being moderated by water instead of graphite and fluoride salt, and sensitivity magnitude effects driven by the amount of FLiBe inserted and its location. New experiments with refined designs to increase the volume and importance of salt in the system could provide improved data. Results also demonstrate that salt systems using thorium and/or 233U fuel will require additional experiments with relevant driver fuels. Increased contributions from the graphite moderator and 19F and covariance between reactions such as 233U fission and 233U radiative capture indicate higher uncertainty contributions from the salt and significant differences in contributions from the fuel species due to underlying uncertainties in their nuclear data.

Neutronic experiments with fluorine rich compounds at LR-0 reactor

Research on molten salt reactor (MSR) neutronics continues in Research Centre Rez (Czech Republic) with experimental work being conducted using fluoride salt that was originally used in the Molten Salt Reactor Experiment (MSRE). Previous results identified significant variations in the neutron spectrum measured in LiF-NaF salt. These variations could originate from the fluorine description in current nuclear data sets. Subsequent experiments were performed to try to confirm this phenomenon. Therefore, another fluorine-rich compound, Teflon, was used for testing. Critical experiments showed slight discrepancies in C/E-1 for both compounds, Teflon and FLIBE, and systematic overestimation of criticality was observed in calculations. Different nuclear data libraries were used for data set testing. For Teflon, the overestimation is higher when using JENDL-3.3, JENDL-4, and RUSFOND-2010 libraries, all three of which share the same inelastic-to-elastic scattering cross section ratio. Calculations using other libraries (ENDF/B-VII.1, ENDF/B-VII.0, JEFF-3.2, JEFF-3.1, and CENDL-3.1) tend to be closer to the experimental value. Neutron spectrum measurement in both substances revealed structure similar to that seen in previous measurements using LiF-NaF salt, which indicates that the neutron spectrum seems to be strongly shaped by fluorine. Discrepancies between experimental and calculational results seem to be larger in the neutron energy range of 100–1300 keV than in higher energies. In the case of neutron spectrum calculation, none of the tested libraries gives overall better results than the others.

Development of the Tritium Transport Analysis Code for the Thorium-Based Molten Salt Reactor

The Thorium-Based Molten Salt Reactor (TMSR) has been highlighted for its safety, economy, and nuclear nonproliferation. A program for developing the TMSR system has been launched in Shanghai Institute of Applied Physics, Chinese Academy of Sciences. In the TMSR system, mixtures of LiF and BeF2, termed FLiBe, are proposed and used as the primary coolant salt, in which tritium is produced mainly by the neutron reactions of lithium. In the TMSR system, at high temperatures, tritium can permeate through metal walls to the surroundings, leading to a potential radiological hazard. Thus, tritium control becomes a major problem hindering the development of the TMSR system. Evaluation of the tritium distribution is necessary for tritium control in the TMSR system. In this study, the Tritium Transport Analysis Code (TTAC) has been developed for simulating the tritium behaviors in the TMSR system (hence, the code TMSR-TTAC), such as tritium chemical forms in coolant salts, tritium transport behaviors, and tritium distribution in the system. The model code is developed by the MATLAB/SIMULINK package, and it is based on the mass balance equations of the tritium-containing species and hydrogen. TMSR-TTAC is benchmarked with the molten salt reactor model, which is based on Molten Salt Reactor Experiment designs. The results show that TMSR-TTAC has the ability to calculate the tritium distribution in the TMSR system.