Modeling the Molten Salt Reactor Experiment With the GOTHIC Code

2020 ◽  
Author(s):  
Rodney Harvill ◽  
Jeff Lane ◽  
John Link ◽  
Anita Gates ◽  
Tom Kindred
Keyword(s):  
Measured Data ◽  
Reactor Power ◽  
Molten Salt Reactor ◽  
Delayed Neutron ◽  
Control Rod ◽  
Decay Heat ◽  
The Core ◽  
Thermal Output ◽  
Fuel Salt ◽  
Good Agreement

Abstract GOTHIC 8.3(QA) includes capabilities for modeling advanced, non-light water cooled reactors. Important capabilities introduced in GOTHIC 8.3(QA) include fluid property tables for various molten salts, an enhancement to the tracer tracking module to allow radioactive decay energy to be released locally in the carrier fluid and other improvements to the neutron kinetics module. With these new capabilities in place, GOTHIC is used to benchmark steady-state and transient conditions in the Molten Salt Reactor Experiment (MSRE), which operated at Oak Ridge National Laboratory from 1965 to 1969. In this experimental reactor, UF4 fuel was dissolved in molten fluoride salt, and criticality could be achieved only in the graphite moderated core. An air-cooled radiator transferred fission and decay heat to the environment. The design thermal output of the MSRE was 10 MWt, but the radiator design limited the output to 8 MWt. The original design parameters neglected the impact of decay heat on system temperatures. GOTHIC is used to benchmark system operating parameters at both the 10 MWt design condition and the 8 MWt operating condition, both with and without decay heat. The cases that include decay heat apply 7% of the nominal thermal output using the eleven decay heat precursors from ASB 9-2 as tracers. The results of the benchmark exhibit good agreement with design and operating data and demonstrate heat-up due to decay heat in the fuel salt outside the core. In the MSRE, delayed neutron precursors are not confined to the core because the fuel and fission products flow through the system. As a result, there are different values for (effective) delayed neutron fraction with and without flow, and the decay of delayed neutron precursors outside the core under full-flow conditions reduces reactivity by 0.212 % δk/k. Zero power physics testing included fuel salt pump start-up and coast-down transients with a control rod automatically moving to maintain criticality. The control rod motion calculated by GOTHIC is a reasonable match to measured data from these transients. Low power testing included a natural convection transient with no control rod motion such that reactor power was responding to heat load demand from the radiator. The reactor power and fuel salt and coolant salt temperatures calculated by GOTHIC exhibit good agreement with measured data.

Numerical Research on Steady Coupling of Neutronics and Thermal-Hydraulics for a Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Changliang Liu ◽  
Libo Qian ◽  
Guanghui Su ◽  
Suizheng Qiu
Keyword(s):  
Thermal Neutron ◽  
Molten Salt ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
Steady Condition ◽  
Delayed Neutron ◽  
The Core ◽  
Production Of Hydrogen ◽  
Fuel Salt ◽  
Salt Flow

The Molten Salt Reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for production of electricity, actinide burning, production of hydrogen, and production of fissile fuels. In this paper, a single-liquid-fueled MSR was selected for conceptual research. For this MSR, a ternary system of 15%LiF-58%NaF-27%BeF2 was proposed as the reactor fuel solvent, coolant and also moderator with ca. 1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. The fuel salt flow makes the MSR different from the conventional reactors using solid fissile materials, and makes the neutronics and thermal-hydraulic coupled strongly, which plays the important role in the research of reactor safety analysis. Therefore, it’s necessary to study the coupling of neutronics and thermal-hydraulic. The theoretical models of neutronics and thermal-hydraulics under steady condition were conducted and calculated by numerical method in this paper. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering flow effect. The thermal-hydraulic model was founded on the base of the fundamental conservation laws: the mass, momentum and energy conservation equations. These two models were coupled through the temperature and heat source. The spatial discretization of the above models is based on the finite volume method (FVM), and the thermal-hydraulic equations are computed by SIMPLER algorithm with domain extension method on the staggered grid system. The distribution of neutron fluxes, the distribution of the temperature and velocity and the distribution of the delayed neutron precursors in the core were obtained. The numerical calculated results show that, the fuel salt flow has little effect to the distribution of fast and thermal neutron fluxes and effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition, and the flow could remove the heat generated by the neutron reactions easily to ensure the reactor safe. The obtained results serve some valuable information for the research and design of this new generation reactor.

scholarly journals Characteristics of a gas-cooled fast reactor with minor actinide loading

2021 ◽  
Vol 8 (2) ◽  
pp. 1-9
Author(s):  
Hoai Nam Tran ◽  
Yasuyoshi Kato ◽  
Van Khanh Hoang ◽  
Sy Minh Tuan Hoang
Keyword(s):  
Fast Reactor ◽  
Supercritical Co2 ◽  
Cycle Length ◽  
Minor Actinide ◽  
Co2 Gas ◽  
Control Rod ◽  
The Core ◽  
Thermal Output ◽  
Long Life ◽  
Gas Turbine Cycle

This paper presents the neutronics characteristics of a prototype gas-cooled (supercritical CO2-cooled) fast reactor (GCFR) with minor actinide (MA) loading in the fuel. The GCFR core is designed with a thermal output of 600 MWt as a part of a direct supercritical CO2 (S-CO2) gas turbine cycle. Transmutation of MAs in the GCFR has been investigated for attaining low burnup reactivity swing and reducing long-life radioactive waste. Minor actinides are loaded uniformly in the fuel regions of the core. The burnup reactivity swing is minimized to 0.11% ∆k/kk’ over the cycle length of 10 years when the MA content is 6.0 wt%. The low burnup reactivity swing enables minimization of control rod operation during burnup. The MA transmutation rate is 42.2 kg/yr, which is equivalent to the production rates in 7 LWRs of the same electrical output.

The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core

2010 ◽  
Author(s):  
Minggang Lang ◽  
Yujie Dong
Keyword(s):  
Gas Turbine ◽  
Cooling System ◽  
Negative Temperature ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
System Pressure ◽  
Temperature Limitation

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basic accidents and beyond design basis accidents of HTR-10GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted and the reactor power increased and the temperature of the core increased. When the neutron flux of power measuring range exceeded 123% and the core outlet temperature was greater than 800°C, the reactor should scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure increased but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1242.4°C. It was a little higher than 1230°C–the fuel temperature limitation of HTR-10. Now the sphere fuel used in HTR-10GT will also be used in HTR-PM and the temperature limitation raised to 1620°C, so the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10. The analysis will be helpful to HTR-PM because they have the same outlet temperature of the core.

Steady Thermal Hydraulics Analysis for a Graphite-Moderated Channel Type Molten Salt Reactor

2008 ◽  
Author(s):  
Dalin Zhang ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Flow Distribution ◽  
Axial Direction ◽  
Theoretical Models ◽  
Outer Wall ◽  
Molten Salt Reactor ◽  
Thermal Hydraulics ◽  
The Core ◽  
Fuel Salt ◽  
Channel Type

The Molten Salt Reactor (MSR) is one of the six GENIV systems capable of breading and burning. In this paper, a graphite-moderated channel type MSR was selected for conceptual research. For this MSR, a ternary system of 0.15LiF-0.58NaF-0.27BeF2 was proposed as the reactor fuel solvent, coolant and also moderator simultaneously with ca.1 mol% UF4 dissolving in it, which circulates through the whole primary loop accompanying fission reaction only in the core. 169 hexagonal graphite elements, each with a central fuel channel, are arranged in the core symmetrically by 30° angles. The theoretical models of the thermal hydraulics under steady condition are conducted in one-twelfth of the core and calculated by the numerical method. The DRAGON code is adopted to calculate the axial and radial power factors. The flow and heat transfer models in the fuel salt and graphite are founded basing on the fundamental mass, momentum and energy equations. The calculated results show the detailed mass flow distribution in the core; and the temperature of the fuel salt, inner and outer wall in the calculated elements along the axial direction are also obtained.

The ATWS Analysis of One Control Rod Withdrawal Out of the HTR-10GT Core Under the Loss of the System Pressure

2014 ◽  
Author(s):  
Minggang Lang
Keyword(s):  
Gas Turbine ◽  
Cooling System ◽  
Negative Temperature ◽  
Reactor Power ◽  
Outlet Temperature ◽  
Fuel Temperature ◽  
Control Rod ◽  
The Core ◽  
Design Basis ◽  
System Pressure

The 10MW High Temperature Gas Cooled Test Reactor (HTR-10) has been built in Institute of Nuclear and New Energy Technology (INET) and has been operating successfully since the beginning of 2003. The core outlet temperature of HTR-10 is 700°C. To verify the technology of gas-turbine direct cycle, at first INET had a plan to increase its core outlet temperature to 750°C and to use a helium gas turbine instead of the steam generator (then the reactor is called HTR-10GT). Though HTR-10 has good intrinsic safety, the design basis accidents and beyond design basis accidents of HTR10-GT must be analyzed according to China’s nuclear regulations due to changed operation parameters. THERMIX code system is used to study the ATWS accident of one control rod withdrawal out of the core by a mistake under the loss of the system pressure. After a control rod in the side reflector was withdrawn out at a speed of 1 cm/s by a mistake, a positive reactivity was inserted. At the same time, the system pressure was supposed to lose by some reason. Thus the reactor power increased and the temperature of the core increased. And the protection system warns with two scram signal: too high of the negative varying rate of the system pressure and too high of the reactor power, which should induce the reactor to scram. It was supposed that all the control rods in the reflectors had been blocked and the reactor could not scram. Thus the accident went on and the core temperature and the system pressure continued to increase but the reactor shutdown at last because of its natural negative temperature reactivity feedback mechanism. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1203.4°C. It was a little bit lower than 1230°C — the fuel temperature limitation of HTR-10 and there is no release of any radioactivity. So the HTR-10GT is safe during the ATWS of one control rod withdrawal out of the core. The paper also compares the analysis result of HTR10-GT to those of HTR-10. The results shows that the HTR-10GT is still safe during the accident though its operating temperature is higher than HTR-10.

Burn-Up Dependency of Control Rod Position at Zero-Power Criticality in the High-Temperature Engineering Test Reactor

2016 ◽  
Vol 3 (1) ◽  
Author(s):  
Yuki Honda ◽  
Nozomu Fujimoto ◽  
Hiroaki Sawahata ◽  
Shoji Takada ◽  
Kazuhiro Sawa
Keyword(s):  
High Temperature ◽  
Temperature Distribution ◽  
Core Temperature ◽  
Measured Data ◽  
Block Type ◽  
Control Rod ◽  
The Core ◽  
Burnable Absorber ◽  
Test Reactor ◽  
Operation Cycle

The high-temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR), which was constructed in Japan. The operating data of HTTR with burn-up to about 370 EFPD (effective full-power days), which are very important for the development of HTGRs, have been collected in both zero-power and powered operations. In the aspects of code validation, the detailed prediction of temperature distribution in the core makes it difficult to validate the calculation code because of difficulty in measuring the core temperature directly in powered operation of the HTTR. In this study, the measured data of the control rod position, while keeping the temperature distribution in the core uniform at criticality in zero-power operation at the beginning of each operation cycle were compared with the calculated results by core physics design code of the HTTR. The measured data of the control rod position were modified based on the core temperature correlation. At the beginning of burn-up, the trends of burn-up characteristics are slightly different between experimental and calculation data. However, the calculated result shows less than 50 mm of small difference (total length of control rod is 4060 mm) to the measured one, which indicates that the calculated results appropriately reproduced burn-up characteristics, such as a decrease in uranium-235, accumulation in plutonium, and decrease in burnable absorber.

Reactor Controllability of 3-Region-Core Molten Salt Reactor System: A Study on Load Following Capability

2006 ◽  
Author(s):  
Takahisa Yamamoto ◽  
Koshi Mitachi ◽  
Masatoshi Nishio
Keyword(s):  
Molten Salt ◽  
Solid Fuel ◽  
Molten Salt Reactor ◽  
Delayed Neutron ◽  
Reactor Kinetics ◽  
External Loop ◽  
Fuel Salt ◽  
Salt Flow ◽  
Point Reactor Kinetics ◽  
Graphite Moderator

The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of ssile fuels (breeding). Thorium (Th) and uranium-233 (233U) are fertile and ssile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development [1]. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the ssion reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. [2] and Shimazu et al. [3] developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow.

scholarly journals An improved control rod selection algorithm for core power control at TRIGA PUSPATI Reactor

2020 ◽  
Vol 14 (1) ◽  
pp. 6362-6379 ◽  
Author(s):  
Mohd Sabri Minhat ◽  
Nurul Adilla Mohd Subha ◽  
Fazilah Hassan ◽  
Norjulia Mohamad Nordin
Keyword(s):  
Power Control ◽  
Reactor Core ◽  
Reactor Power ◽  
Fine Tuning ◽  
Selection Algorithm ◽  
Acceptable Error ◽  
Control Rod ◽  
The Core ◽  
Significant Difference ◽  
Point Kinetics

The 1 MWth TRIGA PUSPATI Reactor known as RTP undergoes more than 37 years of operation in Malaysia. The current core power control utilized Feedback Control Algorithm (FCA) and a conventional Control Rod Selection Algorithm (CRSA). However, the current power tracking performance suffers and increase the workload on Control Rod Drive Mechanism (CRDM) if the range between minimum and maximum rod worth value for each control rod has a significant difference. Thus, it is requiring much time to keep the core power stable at the power demand value within the acceptable error bands for the safety requirement of the RTP. In conventional CRSA, regardless of the rod worth value, the lowest position of the control rod is selected for up-movement to regulate the reactor power with 2% chattering error. To improve this method, a new CRSA is introduced named Single Control Absorbing Rod (SCAR). In SCAR, only one rod with highest reactivity worth value will be selected for coast tuning during transient and the lowest reactivity worth value will be selected for fine-tuning rod movement during steady-state. The simulation model of the reactor core is represented based on point kinetics model, thermal-hydraulic models and reactivity model. The conventional CRSA model included with control rod position dynamic model and actual reactivity worth curve data from RTP. The FCA controller is designed based on Proportional-Integral (PI) controller using MATLAB Simulink simulation. The core power control system is represented by the integration of a reactor core model, CRSA model and FCA controller. To manifest the effectiveness of the proposed SCAR algorithm, the results are compared to the conventional CRSA in both simulation and experimentation. Overall, the results shows that the SCAR algorithm offers generally better results than the conventional CRSA with the reduction in rising time up to 44%, workload up to 35%, settling time up to 26% and chattering error up to 18% of the nominal value.

Stability Validation Based on the OECD/NEA Oskarshamn-2 BWR Benchmark With Toshiba TRAC Version

2016 ◽  
Author(s):  
Takanori Fukunaga ◽  
Noriyuki Yoshida ◽  
Akira Mototani
Keyword(s):  
Initial Conditions ◽  
Safety Analysis ◽  
High Accuracy ◽  
Measured Data ◽  
Regulatory Body ◽  
Neutron Kinetics ◽  
The Core ◽  
Power Oscillations ◽  
Advanced Version ◽  
Good Agreement

We have developed a best-estimate transient code Toshiba TRAC version, a Toshiba advanced version of the TRAC code for BWR. This code is expected to be capable of simulating thermal hydraulic and neutron kinetics phenomena three-dimensionally with high accuracy for BWR transients. A licensing topical report for this code’s validation analyses will be submitted to the Japanese regulatory body for application to future licensing safety analysis. As one of the validation analyses, analysis for BWR stability is also important, and this code’s validation analyses for BWR stability based on the international OECD/NEA Oskarshamn-2 BWR benchmark have been implemented. The Oskarshamn-2 BWR experienced a core instability event, which culminated in diverging power oscillations with a decay ratio greater than 1.3 on February 25, 1999. This paper shows this code’s validation analyses of the core instability event. In conclusion, the initial conditions and the results of the instability event are also in good agreement with the measured data. This code is capable of simulating instability events well.

Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor

2009 ◽  
Author(s):  
Peng Wang ◽  
Suizheng Qiu
Keyword(s):  
Molten Salt ◽  
Flow Velocity ◽  
Design Theory ◽  
Primary Circuit ◽  
Molten Salt Reactor ◽  
Hydraulic Characteristics ◽  
Delayed Neutron ◽  
Delayed Neutrons ◽  
Flow Effect ◽  
Fuel Salt

The new concept Molten Salt Reactor is the only liquid-fuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutrons precursors are drifted by the fuel flow, which leads the spread of delayed neutrons distribution to non-core parts of the primary circuit, and it also can result in a reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the Molten Salt Reactor is quite different from the conventional nuclear reactors using solid fissile materials, and no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particle considering the flow effect of the fuel salt in the Molten Salt Reactor, while the thermal-hydraulic model uses the fundamental conservation laws: the mass, momentum and energy conservation equations. Then the neutronic and thermal-hydraulic calculations were coupled and the influences of inflow temperature and flow velocity on the reactor physical properties were obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor keff. While the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron flux and delayed neutron precursors slightly, and makes significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady state, but affects the delayed neutron precursors’ distribution significantly.

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