High Temperature Reactor System Research Articles

Verification and Demonstration of One-Dimensional Freezing Model in SAM for Salt-Cooled Reactor Analysis Applications

This work presents the development and implementation of the one-dimensional freezing model in system analysis code SAM (System Analysis Module), code verification using analytical solutions, and code demonstration of a postulated overcooling transient for a fluoride salt–cooled high-temperature reactor (FHR) system and safety analysis applications. This paper first summarizes the freezing model, finite element numerical method, and special numerical treatment for handling transitions between single- and two-phase conditions. Analytical solutions are derived for two cases, with and without solid walls, for code verification purposes. As expected, the numerical results predicted by SAM agree very well with the analytical solution. A code demonstration is then performed on a postulated protected overcooling transient event of a generic reference pebble bed FHR design. The code was found to successfully predict salt freezing during such a postulated event. However, due to the lack of salt freezing testing data, code validation is not performed in this work, but will be pursued in future studies when such data become available.

High temperature mechanical properties of diffusion welded alloy 800H

High temperature mechanical properties of diffusion welded Alloy 800H was investigated to fabricate a printed circuit heat exchanger (PCHE) for high temperature reactor (HTR) systems. Surface treatment was performed on Alloy 800H to transform the solubility product (Ksp) to exceed the reaction quotient (Q). The surface treatment facilitated the grain boundary migration across the interface. The yield strengths exceeded the values described in ASME Section III Division 5 Table HBB-I-14.5 up to 760 °C, while the tensile strengths were comparable to Table HBB-3225-1 up to 700 °C. At 760 °C, the tensile strength was ∼30 MPa lower than the code. The stress-to-rupture values exceeded the expected minimum stress-to-rupture values of Alloy 800H described in Table HBB-I-14.6C. The ductility of the diffusion weldment acquired from the tensile test was comparable to the as-received Alloy 800H. Meanwhile, the formation of the secondary precipitates on the interface during the stress-to-rupture test deteriorated the ductility of the diffusion weldment, which induced intergranular fracture.

Synergistic Effect Between Empty Fruit Bunch and HDPE on Products Yield and Functional Group Content in Bio-fuels Produced from Co-liquefaction Process Under Supercritical Methanol Condition

Synergism between biomass and plastic waste during co-processing reported to produce higher products yield with enhanced quality which thereby solving the problems of high oxygenated compounds and low yields with high viscosity in oil from its individual processing. However, study on synergistic effect (SE) between these feedstocks is considered lacking. In this study, the SE of co-liquefying empty fruit bunch (EFB) and high-density polyethylene (HDPE) in sc-MeOH was investigated using a high-pressure, high-temperature batch-wise reactor system at 260 °C for 30 minutes and at various EFB-HDPE ratios to determine the best ratio for bio-fuel yield and its functional group content. It was found that 25% EFB:75% HDPE (wt/wt %) provides the most significant SE (≈ 8 wt%) for bio-fuel yield and the most considerable reduction in solid residue formation under the studied conditions. The presence of hydroxyl groups from EFB and aliphatic hydrocarbons from HDPE, along with new carbonyl compounds, further highlights the SE resulting from the combined processing of these materials. The findings suggest that optimizing the ratio of biomass to plastic waste in co-liquefaction can significantly enhance the efficiency and yield of bio-fuels while minimizing waste, providing a promising approach for sustainable energy production and waste management.

850℃ 고온에서 Alloy 800H 모재와 용접재의 저사이클 피로 거동

Alloy 800H is one of the candidate materials for key structural components such as control rod systems, hot gas duct, core barrels, core supports and shut-down cooling systems for the very high temperature reactor (VHTR) system. The aim of this paper is to compare the low-cycle fatigue behaviors of the alloy 800H base metal and weldments at a high temperature of 850℃. Low cycle fatigue tests were carried out for four different total strain ranges of 0.6, 0.9, 1.2 and 1.5% under constant strain rate of 10-3/s. In all test conditions, the fatigue life of base metal and weldments decreased inversely proportional to the total strain range. The cyclic stress response behavior showed cyclic softening phenomenon for both materials. The low cycle fatigue life for base metal and weldments were predicted by the Coffin-Manson-Basquin law. The results showed that transition life of the weldments is shorter than that of the base metal.

VALIDATION OF WIMS11 FOR SMALL MODULAR REACTOR ANALYSIS

The WIMS (Winfrith Improved Multigroup Scheme) reactor physics code is actively being developed for whole core modelling of a range of Small Modular Reactor types including the Pressurized Water Reactor (PWR), High Temperature Reactor (HTR), and Liquid Metal Cooled Fast reactor (LMFR). These developments include the capability for whole core multiphysics modelling with neutronics and thermal hydraulic feedback, as well as methods to determine the power deposition from neutron and gamma heating. Flux solutions are obtained using a wide variety of deterministic methods including diffusion theory, SP3, and full transport with the method of characteristics and Sn discrete ordinates methods, as well as multi-group Monte Carlo methods. The SP3 method allows both steady state and time dependent transient solutions by solving the time dependent SP3 equations. A wide variety of nuclear data libraries are available with WIMS including data from the JEF3.3, ENDF/B-VII.0 and CENDL3.1 nuclear data evaluations. This paper presents validation of the latest version of the WIMS code, WIMS11, for PWR and HTR systems. Comparisons are made against physics data obtained from the OECD/NEA PWR Watts Bar multi-physics benchmark and the IAEA HTR-10 benchmark, as well as neutron and gamma heating experiments that took place on the NESSUS reactor at Winfrith in the United Kingdom. In each case, validation of WIMS has been obtained by comparison either against measured data, or results provided by other benchmark participants that have been obtained with alternative deterministic or Monte Carlo methods.

The Fuel Cycle Implications of Nuclear Process Heat

International and UK fuel cycle scenario analyses performed to date have focused on nuclear plants producing electricity without considering in detail the other potential drivers for nuclear power, such as industrial process heat. Part of the reason behind the restricted applications of nuclear power is because the assumptions behind the future scenario are not fully captured, for example how big are demands from different sectors? Here we present a means to fully capture the potential opportunities for nuclear power using Sankey diagrams and then, using this information, consider for the first time in the UK the fuel cycle implications of decarbonising industrial heat demand in the year 2050 with nuclear power using the ORION fuel cycle code to study attributes related to spent fuel, uranium demand and decay heat from the spent fuel. We show that even in high industrial energy demand scenarios, the sensitivity of spent fuel masses and decay heat to the types of reactor deployed is relatively small compared to the greater fuel cycle demands from large-scale deployment of nuclear plants for electricity production. However, the sensitivity of spent fuel volumes depends heavily on the extent to which High Temperature Reactor and Light Water Reactor systems operating on a once-through cycle are deployed.

Improved Performance and Efficiency of Lanthanum–Strontium–Manganese Perovskites Undergoing Isothermal Redox Cycling under Controlled pH2O/pH2

Two-step thermochemical redox cycles represent an attractive pathway toward producing solar fuels via H2O and CO2 splitting. These cycles typically operate at two different temperatures, but isothermal cycling has been proposed as a means to minimize the sensible heating requirements of the solid phase. Recently, a theoretical analysis showed that lanthanum–strontium–manganese-based perovskites compare favorably to the state-of-the-art material, ceria, when evaluated in the context of isothermal cycling. In this work, a high-temperature reactor system that affords precise control of partial pressures of H2O and H2 (pH2O/pH2), and thus reduction and oxidation conditions, was utilized for isothermal cycling of ceria, La0.65Sr0.35MnO3 (LSM35) and La0.6Sr0.4MnO3 (LSM40), in order to validate theoretical predictions experimentally. Data are measured between 1300 and 1400 °C and between oxygen partial pressures (pO2) of 8 × 10–7 atm (reduction, pH2O/pH2 in the range of 62.73–203.17) and 5 × 10–5 atm (oxidation, pH2O/pH2 in the range of 250.92–507.93). Both LSM35 and LSM40 significantly outperformed ceria while operating isothermally in terms of H2 production and performance (Π), a new metric introduced in this article. This material-specific metric is related to efficiency, but enables straightforward comparison of materials directly from only a few measured experimental data points. Overall, experimental data agree well with theoretical predictions, giving promise to isothermal operation for this class of materials. Overall, this work provides motivation for material exploration and development of materials that undergo large changes in nonstoichiometry in the pO2 range between 8 × 10–7 atm and 5 × 10–5 atm. Importantly, because operation is isothermal or near-isothermal, material exploration is not limited to materials possessing large entropic changes similar to those aimed specifically at temperature-swing redox cycling, opening the door to new possibilities and different material classes from those have been considered prior.

ASME Boiler and Pressure Vessel Code Roadmap for Compact Heat Exchangers in High Temperature Reactors

Abstract The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHXs) for the next generation of nuclear reactors. The DOE is sponsoring research to support the development and deployment of CHXs for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) that includes university research institutes, national laboratories, manufacturers, and industry experts. The objective is to enable the use of CHX designs in advanced reactor service. A necessary step for achieving this objective is to ensure that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a CHX in an advanced reactor. A strategy for ASME Boiler and Pressure Vessel Code, Section XI, Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will also be required. The purpose of this ASME Code Roadmap is to identify the research gaps impeding the development of suitable construction and ISI rules for CHXs for high temperature reactor service and to provide a framework to utilize the research project results consistent with the expectations and needs of the industry and future owners.

Multi-equipment maintenance decision-making method for high temperature reactor based on maintenance event correlation

The composition of the high temperature reactor system is complex, effective maintenance management strategy can reduce the number of shutdowns and reduce maintenance costs during the operation of the equipment and caused by deterioration, improve system reliability, and also avoid system failures, maintain or restore the system to the most appropriate standard of reliability, availability and safety. This paper proposes a methodology on maintenance decision making with target of minimal downtime by analyse the associated relationships of different maintenance activities to realize merge maintenance.

Maintenance Expense Optimization Method for High Temperature Reactor Equipment Group

The composition of the high-temperature reactor system is relatively complicated. The failure mechanism and degradation process of each component in the system are not the same, and the timing of failure is not the same, resulting it impossible to ensure all components being maintained by preventive planning maintenance. According to the fault correlation, time correlation, functional correlation and structural correlation between components in the multi-component system, the components that meet the maintenance conditions within a certain period of time can be integrated and repaired, which can not only share the fixed cost, but also reduce the downtime. It also increases the overall availability of the equipment.

Uniaxial Low-Cycle Fatigue Study of Alloy 800H Weldments at 700 °C

Alloy 800H is currently under consideration to be used in a very high temperature reactor system. This paper describes the experimental low-cycle fatigue study on Alloy 800H base metal and weldments in the air with fully-reversed strain control at 700 °C. The recommended modern Alloy 800H weldments were fabricated by gas tungsten arc welding. The observed fatigue life is found to be comparable between base metal and weldments at this temperature, indicating the welding technique can compete with base material. The influence of hold time was further investigated at the maximum tensile strain of 0.6% total strain range condition for 60 s, and it is found to reduce the fatigue life. For most of the tests, the base metal revealed an initial cyclic hardening stress response, otherwise, the weldment was entirely softened with cycles. The addition of hold time is also found to induce an unequal cyclic stress response in both materials. Accordingly, for continuous low-cycle fatigue tests, fatigue cracks were initiated and propagated in a classical transgranular fracture mode. However, the hold time affected the mixed-mode fracture mode by means of a transition from transgranular to intergranular fracture.

Analysis of different nuclear data libraries on the Fluoride-salt-cooled high-temperature reactor

This work presents a comparison of Fluoride-salt-cool High-temperature Reactor (FHR) system calculations using the ENDF/B-VII.1, JEFF-3.2, JEFF-3.3 and ENDF/B-VIII.0 nuclear data libraries. Analysis of a simplified model of the Shanghai Institute of Applied Physics TMSR-SF1 reactor using both Serpent and MCNP was performed. Unit cell and full core models based on a specification developed at the University of California, Berkeley were used in this work. The use of the newer cross section libraries produced an increase of thermal fission of 235U by 0.5–1%. The reactivity of the system in a fresh configuration was increased by 270–600 pcm, thermal flux was reduced by ∼1% and the effect on reactivity of the system as a function of 7Li purity of salt coolant was also reduced by ∼2%. Neutron leakage, reaction rates, neutron flux spectra, thermal flux distributions and the effect of spatially homogenising regions of the system were also calculated and analysed.

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

The release of fission products from coated particle fuel to primary coolant, as well as the activation of coolant and impurities, were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system, and the activity of radionuclides accumulated in the coolant during normal operation was calculated. The release rate (release fraction per unit time) of fission products was calculated with STACY code, which is modelled mainly based on the Fick’s law, while the activation of coolant and impurities was calculated with SCALE code. The accumulation of radionuclides in the coolant has been calculated with a simplified model, which is generally a time integration considering the generation and decay of radionuclides. The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown. In operation condition, health-impacts related nuclides such as 3H, and 14C originate from the activation of lithium and coolant impurities including carbon, nitrogen, and oxygen. According to the calculated effective cross sections of neutron activation, 6Li and 14N are the dominant 3H production source and 14C production source, respectively. Considering the high production rate, 3H and 14C should be treated before being released to the environment.

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

Abstract Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (Tco) and the hot outlet temperature of the intermediate heat exchanger (Tho2) by manipulating the hot-side flow rates of the heat exchangers (Fh/Fh2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (Tco) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

A study on transmutation of LLFPs using various types of HTGRs

In order to investigate the potential of high temperature gas-cooled reactors (HTGRs) for transmutation of long-lived fission products (LLFPs), numerical simulation of four types of HTGRs were carried out. In addition to the gas-turbine high temperature reactor system “GTHTR300”, which is the subject of our previous research, a small modular HTGR plant “HTR50S” and two types of plutonium burner HTGRs “Clean Burn with MA” and “Clean Burn without MA” were considered. The simulation results show that an early realization of LLFP transmutation using a compact HTGR may be possible since the HTR50S can transmute fair amount of LLFPs for its thermal output. The Clean Burn with MA can transmute a limited amount of LLFPs. However, an efficient LLFP transmutation using the Clean Burn without MA seems to be convincing as it is able to achieve very high burn-ups and produce LLFP transmutation more than GTHTR300. Based on these results, we propose utilization of variety of HTGRs for LLFP transmutation and storage.

Thermal Analysis of Spent Nuclear Fuels Repository

In the first part, Pressurized Water Reactor (PWR), Very High-Temperature Reactor (VHTR) and Accelerator-Driven Subcritical Reactor System (ADS) spent fuels (SF) were evaluated to the thermal of the spent fuel pool (SFP) without an external cooling system. The goal is to compare the water boiling time of the pool storing different types of spent nuclear fuels. This study used the software Ansys Workbench 16.2 - student version. For the VHTR, two types of fuel were analyzed: (Th,TRU)O2 and UO2. This part of the studies were performed for wet storage condition using a single type of SF and decay heat values at times t=0 and t=10 years after the reactor discharge. The Ansys CFX module was used and the results show that the time that water takes to reach the boiling point varies from 2.4 minutes for the case of VHTR-(Th,TRU)O2 SF at time t=0 year after reactor discharge until 32.4 hours for the case of PWR SF at time t=10 years after the discharge reactor.The second part of this work consists of modeling a geological repository. Firstly, the temperature evaluation of the spent fuel from a PWR was analyzed. A PWR canister was simulated using the Ansys transient thermal module. Then the temperature of canister could be computed during the time spent on a portion of a geological repository. The mean temperature on the canister surface increased during the first nine years, reaching a plateau at 35.5°C between the tenth and twentieth years after the geological disposal. The idea is to extend this study for the other systems analyzed in the first part. The idea is to include in the study, the spent fuels from VHTR and ADS and to compare the canister behavior using different spent fuels.

Development of Forming Processes for MoNiCr Alloy

Nickel based super alloy MoNiCr was developed as the material with high corrosion and creep resistance against aggressive molten fluoride salt environment. This medium is used by Fluoride salt cooled High temperature Reactor (FHR) and Molten Salt Reactor (MSR) systems. The elements Mo, Cr and Fe are most important alloying elements accompanied by very low amount of Al + Ti. Necessity to prepare hot formed semi products such as sheets, bars, wires and tubes was the motivation for this new experimental program execution searching for suitable hot forming conditions. In the case of MoNiCr components forming, recrystallization is a crucial process allowing successful forming. Special attention was paid to this process after finding out, that this process is running very slowly or even not running at all, if inappropriate conditions are met. Proper temperature and strain rate range were determined and furthermore the influence of cold deformation on recrystallization progress was determined. It was found that cold deformation before hot forming as well as after hot forming can be successfully implemented into technological chain. Almost fully recrystallized microstructure can be achieved using synergy of both cold and hot deformation.

Comparison of fast neutron spectra in graphite and FLINA salt inserted in well-defined core assembled in LR-0 reactor

The present paper aims to compare the calculated and measured spectra after insertion of candidate materials for the Molten salt reactor/Fluoride cooled high temperature reactor system concept into the LR-0 reactor. The calculation is realized with MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-4, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Additionally, comparisons between the slowing down power of each media were performed. The slowing down properties are important parameters affecting the thickness of moderator media in a reactor.

Alloy 617 모재와 용접부재의 저사이클피로 거동에 관한 실험적 고찰

Alloy 617 is the one of the leading candidate materials for intermediate heat exchangers(IHX) of a very high temperature reactor(VHTR) system. Some of the components are joined by many welding techniques and therefore the welded joints are inevitable in the construction of systems. In the present paper, the low cycle fatigue(LCF) behaviors of Alloy 617 base metal(BM) and the gas tungsten arc welded (GTAWed) weld joints(WJ) are investigated experimentally under strain controlled LCF tests. Fully axial total-strain controlled tests have been conducted at room temperature with total strain ranges of 0.6, 0.9, 1.2 and 1.5%. The weld joints have shown a lower fatigue lives compared with base metals at all the testing conditions. The weld joints have shown a higher cyclic stress response behavior than base metal. Both BM and WJ exhibited cyclic strain hardening behavior, depending on the total strain range. In addition, the strain-life parameters for BM and WJ were determined, based on Coffin-Manson equations.

저사이클피로 하중하의 Alloy 617 모재와 용접부재에 대한 피로 수명 평가

Generally, the mechanical components and structures are joined by many welding techniques, and therefore the welded joints are inevitable in the construction of structures. The Alloy 617 was initially developed for high temperature applications above <TEX>$800^{\circ}C$</TEX>. It is often considered for use in aircraft and gas turbines, chemical manufacturing components, and power generation structures. Especially, the Alloy 617 is the primary candidate for construction of intermediate heat exchanger (IHX) on a very high temperature reactor (VHTR) system. In the present paper, the low cycle fatigue (LCF) life of Alloy 617 base metal (BM) and the gas tungsten arc welded (GTAWed) weld joints (WJ) are evaluated by using the previous experimental results under strain controlled LCF tests. The LCF tests have been performed at room temperature with total strain ranges of 0.6, 0.9, 1.2 and 1.5%. The LCF lives for the BM and WJ have been evaluated from the Coffin-Manson and strain energy based life methods. For both the BM and WJ, the LCF lives predicted by both Coffin-Manson and strain energy based life methods was found to well coincide with the experimental data.