High Temperature Gas-cooled Test Reactor Research Articles

Multiphysics Analysis of PLOFC and DLOFC Accidents in the HTGR-TR

High-temperature, gas-cooled reactors (HTGRs) are promising advanced reactor designs. Leveraging passive safety features, higher core outlet temperatures, previous commercial operational experience, and potential coupling with nonelectrical systems, HTGRs present many advantages over traditional light water reactors and other advanced reactor designs. The High-Temperature, Gas-Cooled Test Reactor (HTGR-TR) is a point design developed by Idaho National Laboratory as a response to the need for the expansion of U.S. test reactor capabilities. The HTGR-TR was designed to serve as a test reactor for Generation-IV small modular prismatic block HTGRs and to provide irradiation data, two crucial contributions to the development of advanced reactors. As in any reactor design, it is necessary to understand the behavior of the reactor during accident scenarios to determine the overall safety of the design. Pressurized loss of forced cooling (PLOFC) and depressurized loss of forced cooling (DLOFC) are two accidents that can occur in HTGRs and that present challenges to decay heat removal. In order to understand the behavior of the fuel during these accidents, the resulting mechanical behavior and fission product migration in the fuel are evaluated. Using the HTGR-TR design, which utilizes tristructural isotropic (TRISO) fuel, a multiphysics analysis during the PLOFC and DLOFC transients was performed. This study utilizes neutronics, thermal hydraulics, and fuel performance modeling to analyze the behavior of TRISO particles during steady-state operation and the described transients. This analysis aims to characterize the hoop stress and strain during the transients, as well as predict fission product migration and radiological release. Due to the startup conditions of the HTGR-TR, the silicon carbide (SiC) layer for both of the transients and steady-state conditions are initially in compression for both the hoop stress and strain. The DLOFC transient at the beginning of cycle was the most challenging overall to the fuel, having the closest to tensile behavior and the closest to tensile strain. Because the integrity of the SiC layer was not challenged during any of the transients, the release of 90Sr and 137Cs was found to be negligible. The PLOFC transient at the end of cycle resulted in the largest 110mAg release and subsequent radioactivity increase. However, the increase in radioactivity was minor when compared to typical operating conditions due to the relatively large diffusivity of 110mAg though SiC at normal operating temperatures. Highlights include Neutronics and thermal-hydraulic modeling generating steady-state and transient conditions for use in fuel performance analysis.2. Detailed BISON analysis demonstrating the excellent performance of TRISO fuel during PLOFC and DLOFC transients.3. Fuel behavior showing negligible fission product release.

Thermal-Hydraulic Analysis of Pebble Bed High Temperature Gas-Cooled Reactor After Shutdown Based on FLUENT Software

In order to provide a reliable tool for thermal-hydraulic simulation of pebble bed high temperature gas-cooled reactors (HTGRs), a two-dimensional model was developed based on the porous media model and user-defined scalar (UDS) function of FLUENT software. Then, the model was applied to the numerical simulation of the shutdown test of the 10 MW high temperature gas-cooled test reactor (HTR-10) at 9 MW power level, and the temperature distribution and flow field distribution in the reactor were obtained and compared with the results of the experimental data. The reliability of the model in this paper was verified. Based on the model, the effects of the water-cooled panel temperature and the initial core temperature on the thermal-hydraulic characteristics of HTR-10 after shutdown were further explored. The results show that there is a decoupling phenomenon between the residual heat transfer within the core and the heat dissipation of the pressure vessel. The initial core temperature has relatively little effect on the heat dissipation and maximum temperature of the pressure vessel, but it has a significant impact on the thermal characteristics of the core area.

Numerical simulation on flow and reaction characteristics for catalytic region in helium-heated steam reformer coupled with HTR-10

In this paper, one dimension steady-state and pseudo-homogeneous model of the hydrogen production reformer in the steam methane reforming (SMR) process using the high temperature gas-cooled test reactor of 10 MW (HTR-10) as heat source (HTR-10/SMR) is simulated based on the semi-coupled solution method using matlab2016b. Firstly, The axial variations of hydrogen production performance including temperature (T2), methane conversion rate (XCH4), carbon monoxide conversion rate (XCO) and hydrogen production per mole methane (YH2/CH4 or total hydrogen production (YtH2)) as well as thermal efficiency (η) are investigated. Then, it is noted that YH2/CH4 decreases from 2.92 mol H2/mol CH4 to 2.53 mol H2/mol CH4. However, YtH2 increases from 2.33 kmol/h to 3.23 kmol/h with the increase of the inlet process gas flow rate (F2inlet) from 4 kmol/h to 6 kmol/h. Finally, YtH2 increases with steam to carbon ratio (S) when S < 4 and decreases when S > 4. The maximum value of YtH2 with 3.03 kmol/h is obtained at S = 4. Especially, XCH4, XCO, YH2/CH4 and η are in the range of 70.54%–90.78%, 38.56%–54.30%, 2.50–3.07 mol/mol CH4 and 54.55%–64.57%, respectively, when T3_inlet is in the range of 850–950 °C. The effect of the process gas temperature and pressure on the system performance can be neglected.

Preliminary study on HTR-10 operating at higher outlet temperature

High Temperature Gas-cooled Reactor (HTGR), which has well-known safety features and high temperature heat supply capability, is expected to be widely used for heat supply and technology heat utilization including the hydrogen production, and so contributing to the reduction of carbon dioxide emissions in various sectors. The 10 MW High Temperature gas-cooled test Reactor (HTR-10) had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. The first criticality of HTR-10 at air condition was realized on December 1, 2000, and the full power operation for 72 h on January 29, 2003.Supported by Chinese National S&T Major Project, HTGR for hydrogen production are now being studied. The physical and thermal hydraulic design to raise the outlet helium temperature of the HTR-10 reactor core from 700 °C to 850–1000 °C is carried out.In this paper, the preliminary thermal hydraulic design of the HTR-10 with the outlet helium temperature of 950 °C (HTR-10H) is introduced. The power density distribution, the fuel temperature distribution and the reactor pressure vessel (RPV) temperature are studied to identify what need to be focused on next. Besides, the DLOFC accident has been studied to evaluate the safety feature of the HTR-10 operating under higher core temperature and outlet temperature.Due to the lower power and power density, as well as the large heat capacity of the graphite reflector and the carbon brick, the maximal fuel temperature during the DLOFC would not exceed the limitation of 1620 °C (250th meeting of the reactor safety commission, 1990). Nevertheless, the higher fuel temperature during the normal operation should be of concern.

Pre-conceptual high temperature gas-cooled microreactor design utilizing two-phase composite moderators. Part I: Microreactor design and reactor performance

In this work, an approach is demonstrated for designing a microreactor based on a High Temperature Gas-Cooled Reactor (HTGR) platform intended for use in remote regions or for emergency response. The pre-conceptual microreactor design generated in this study is based on the Idaho National Laboratory's High Temperature Gas-Cooled Test Reactor (HTGTR) point design. Major design changes to the HTGTR were made to transform the design to a microreactor based on constraints inspired by open literature information from the Office of Secretary of Defense (OSD)/Strategic Capabilities Office (SCO) Pele Program. The primary purpose of this study is to compare reactor performance by evaluating key figures of merit for systems employing two-phase composite moderators (magnesia matrix with either an entrained beryllium or hydride containing moderating phase) relative to a reference graphite moderated case. Design parameters were selected to maximize reactor cycle performance for a given microreactor design with additional performance metrics such as natural resource requirement and fuel cycle costs also calculated. Our results demonstrate that a microreactor design based on the HTGR can be achieved by implementing a high fuel density. By using two-phase composite beryllium- or hydride-based moderators instead of graphite, substantial improvements can be realized in the reactor cycle performance, natural resource utilization, and fuel cycle cost.

Analysis of gamma spectra from burnup measurement in HTR-10

HTR-10, the 10 MW pebble-bed High Temperature Gas-cooled Reactor-Test Module designed by Tsinghua University, achieved its first criticality in 2000. One of the unique features of HTR-10 is the continuous fuel feeding achieved by its fuel cycle system, in which the fuel sphere is allowed to pass through the core several times before discharged out. During the fuel circulation, the on-line burnup measurement system is used to ensure fuel credit. Up to now, after nearly 21 years of operation, huge data including gamma spectra have been collected by this on-line burnup measurement system. In this paper, the gamma spectra data collected from Nov. 2005 to Mar. 2006 are analyzed to identify the radionuclides left in the irradiated fuel sphere after a long decay. The analysis includes energy identification, decay scheme confirmation, activity estimation and so on. This work shows that radionuclides and corresponding activities are preliminarily identified, which lays a base for R&D of new burnup measurement apparatus and the profiling of fuel irradiation history. That would help to increase accuracy for the nuclear material accounting and to improve the operation strategy of pebble bed high-temperature gas-cooled reactors.

Preliminary Study on HTR-10 Operating in Higher Outlet Temperature

High Temperature Gas-cooled Reactor (HTGR), which has well-known safety features and high temperature heat supply capability, is expected to be widely used for heat supply and technology heat utilization including the hydrogen production, and so contributing to the reduction of carbon dioxide emissions in various sectors. The 10 MW High Temperature gas-cooled test Reactor (HTR-10) had been constructed and operated in China as a pilot plant to demonstrate the inherent safety features of the modular HTGR. The first criticality of HTR-10 at air condition was realized on December 1, 2000, and the full power operation for 72 h on January 29, 2003. Supported by Chinese National S&T Major Project, HTGR for hydrogen production are now being studied. The physical and thermal hydraulic design to raise the outlet helium temperature of the HTR-10 reactor core from 700 °C to 850~1000 °C is carried out. In this paper, the preliminary thermal hydraulic design of the HTR- 10 with the outlet helium temperature of 950 °C (HTR-10H) is introduced. The power density distribution, the fuel temperature distribution and the reactor pressure vessel (RPV) temperature are studied to identify what need to be focused on next. Besides, the typical DLOFC accident has been studied to evaluate the safety feature of the HTR-10 operating under higher core temperature and outlet temperature. The preliminary results show that, operated at the higher outlet helium temperature, the original acceptance criteria for HTR-10 will be challenged. In the future, the design optimization, as well as the possible modification of these acceptance criteria, which were set more than two decades ago, should be studied based on the current knowledge of the fuel element properties and structure material properties.

Simulation of the HTR-10 Operation History With the PANGU Code

The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-cooled Reactor (HTGR) in China, which was operated from January 2003 to May 2007. The HTR-10 operation history provides very important data for the validation of HTGR codes. In this paper, the HTR-10 operation history is simulated with the PANGU code, which has been recently developed for HTGR reactor physics analysis and design. Models and parameters are constructed based on the measured data of the actual conditions. The simulation results agree well with the measurements in all steady-state power periods. The discrepancy of keff is generally below 0.5%, and the discrepancy of coolant outlet temperature is generally below 5°C. It is also figured out that the burnup of graphite impurities has considerable influence on the keff at the end of the operation history, which can cause over 1.5% discrepancy when neglecting the burnup of graphite impurities. By this work, the PANGU code’s applicability in actual HTGR fuel cycle simulations is demonstrated.

Summary of Tritium Source Term Study in 10 MW High Temperature Gas-Cooled Test Reactor

Tritium (3H) has been increasingly researched when assessing the environmental impact of nuclear reactors and other nuclear facilities because it is widely present in nuclear systems and can easily enter the environment. The first pebble-bed gas-cooled test reactor in China, the 10 MW high temperature gas-cooled test reactor (HTR-10), uses helium, graphite, and graphite spheres containing embedded tristructural-isotropic–coated particles as primary coolant, reflectors, and fuel elements, respectively. Several experiments that involved the 3H source term in HTR-10 were performed, and they measured the 3H specific activity and its distribution in the irradiated graphite spheres from the core, 3H activity concentration in the primary helium, 3H activity concentration during the regeneration of the molecular sieve adsorber in the helium purification system, and 3H amount in the gaseous effluent discharge from the stack. The experimental data were summarized and compared with the theoretical predictions. The balance diagram of the 3H source term in HTR-10 is introduced in this paper. Sensitivity analysis was performed to illustrate the effect of the 3He abundance in the primary helium and Li content in the graphite reflectors on the 3H activity concentration in the primary coolant of HTR-10. The interactions between graphite and different hydrogen isotopes (1H, 3H, 1H2, 1H3H, and 3H2) were investigated using first-principles calculations and the diffusion theory. The results indicated that molecular 3H tended to diffuse in graphite.

Scaling Analysis of Reactor Cavity Cooling System in HTR

Reactor cavity cooling systems (RCCSs) ensure the physical integrity of the containment structures in a high-temperature gas-cooled test reactor (HTR-10) and a high-temperature gas-cooled pebble-bed module reactor (HTR-PM). HTR-10 is a graphite-moderated and helium-cooled pebble-bed reactor prototype designed to demonstrate the technical feasibility and safety of the pebble-bed reactor design concept under normal and accident conditions. This prototype served as a proof of concept for the HTR-PM that shares several design similarities with the HTR-10, including a reactor cavity that requires cooling owing to the high core outlet temperature. The RCCS conceived in the design of both the reactors increases the inherent safety of the system by dissipating heat through passive heat removal processes. This paper proposes an RCCS model for system-scale analysis. The conventional scale method is adopted to determine the conditions necessary for complete similarity between two RCCSs in the steady-state flow regime. In addition, a scaling evaluation between the RCCSs of both the HTR-10 (model) and HTR-PM (prototype) is performed using the proposed RCCS model based on data from two benchmark problems: pressurized and depressurized loss of forced cooling. This evaluation shows that the RCCSs of the HTR-10 (model) and HTR-PM (prototype) show similarity to a specific operational condition in each of the problems analyzed.

The application of the DEM-based burnup construction method in the pebble burnup history analysis in HTR-10

In the 10 MW high temperature gas-cooled test reactor (HTR-10) operation history, the reactor experienced frequent shutdown periods and irradiation periods with varied power. The complicated operation history is supposed to influence the pebble burnup history. Therefore, this paper firstly introduces a DEM-based pebble burnup tracking method, which combines the DEM pebble flow simulation and the pebble burnup calculation. The method can track the burnup of each pebble in the core individually. Then the method is applied to construct the pebble burnup under the complicated HTR-10 real operation history, and the construction results are analyzed to investigate the influence of operation history on the pebble burnup history. The analyzing results show that the average final pebble burnup at the end of the history is 17–18 GWd/tU. Most of the pebbles experience 2–3 cycles before the reactor finally stops. Moreover, the correlation between the pebble burnup and the Cs-137 activity is found to be influenced by the operation history. The tail of the original single linear distribution turns to be double linear distribution, which brings system error for determining the pebble burnup by Cs-137 activity. In addition, a method that can correct the system error is promised with the help of some recommended auxiliary nuclides.

A one-dimensional code of the passive residual heat removal system for the modular high temperature gas-cooled reactor

The characteristics of the passive residual heat removal system (PRHRS) are very important for the safety of the modular high temperature gas-cooled reactor (MHTGR). In this study, a one-dimensional code used to calculate the transient characteristics of the PRHRS of the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) was developed using the C programming language. Water loop and air cooled condenser are the crucial parts of the PRHRS. Their characteristics under the startup condition, normal operating condition and shutdown condition of HTR-10 were respectively calculated with the one-dimensional code. To validate the code, the computational results were compared with the experimental results obtained from the HTR-10. Also, the flow and heat transfer conditions in the water loop and the air cooled tower were also analyzed. The results show that the computational results are in good agreement with the experimental results. Under the normal operating condition, the characteristics of the PRHRS are mainly affected by the environmental temperature. Under the startup condition and the shutdown condition, the characteristics of the PRHRS are affected by both the nuclear power and the environmental temperature, and the effects of the nuclear power play a dominant role.

Thermal behavior of the HTR-10 under combined PLOFC and ATWS condition initiated by unscrammed control rod withdrawal

Two tests initiated by unscrammed control rod withdrawal were performed on the High Temperature Gas-Cooled Reactor-Test Module (HTR-10) in November 2003 after the reactor achieved its full power, and the test conditions represented a typical transient scenario of modular high-temperature reactors (HTRs), called pressurized loss of forced cooling, and anticipated transient without scram. Based on the test parameters, the HTR-10 thermal behaviors under the test conditions were studied with the help of the system analysis code THERMIX. The combination of the test results and the investigation results makes the HTR-10 safety potential better understood. Key phenomena, such as the helium natural circulation and the temperature redistribution in the reactor, were revealed. As the safety feature of most significance, there is a large margin between the maximum fuel temperature and its safety limit in each test. Temperatures of thermocouples in different components were calculated by THERMIX and compared with the test values. The applicability of the code was verified by good agreement obtained from the comparison.

Three-dimensional numerical investigation of heat transfer process in the reactor cavity of HTR-10

The 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is the first Modular High Temperature Gas-cooled Reactor (MHTGR) in China. The heat transfer process in the reactor cavity of HTR-10, which was very important for the decay heat to be removed, was investigated with a three-dimensional numerical model in this study. The temperature distributions of air, water cooled panel and water in five experimental conditions were analyzed. The heat transfer rates on the reactor pressure vessel (RPV) wall, front and back wall of the water cooled panel, and outer wall of the water cooled pipe were calculated. Besides, the percentages of radiation heat transfer rate on these walls were given, and the variations of them with the RPV temperature rising were also analyzed. The results show that radiation between the RPV and the steel sheet, heat conduction between the steel sheet and the water cooled pipe, and convection between the water cooled pipe and the water form the main way that the heat is removed, especially when the RPV temperature is high. The maximum temperature of air occurs on the top of the RPV. In addition, as the RPV temperature rises, the heat transfer rates and the percentages of radiation all increase.

Dynamic response of the HTR-10 under the control rod withdrawal test without scram

As one of the safety demonstration tests performed on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10), a control rod withdrawal (CRW) test was conducted without reactor scram. In the test process, about 5‰ ∆k/k positive reactivity was introduced into the core at 30% rated power.In this study, the CRW test is simulated using the THERMIX code and the analysis result is compared with the test data. The code reproduces the reactor power well except some discrepancies in predicting the first peak power after the recriticality. Thus, the basic simulation capability of THERMIX can be verified. Besides, the temperature coefficient of reactivity (TCR), the residual heat level (RHL) and the xenon poisoning effect (XPE) are considered as the more crucial aspects affecting the THERMIX simulation accuracy. Therefore, sensitivity studies are performed based on the assumed changes of TCR, RHL and XPE with the objective of guiding further efforts in improving the accuracy of the future post-test simulations.

Numerical simulation on heat transfer process in the reactor cavity of modular high temperature gas-cooled reactor

In order to enhance the safety of the reactor, the passive residual heat removal system (PRHRS) is adopted by both the 10MW High Temperature Gas-cooled Test Reactor (HTR-10) and the High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM). The heat transfer process in the reactor cavity is very important for the whole PRHRS to remove the heat from the reactor, especially once an accident occurs. A three-dimensional numerical model used to simulate the heat transfer process in the reactor cavity of HTR-10 was developed in this study, in which the flow of water in the water cooled pipe was included. The effects of different symmetry boundary conditions on the heat transfer process were studied. Three experimental conditions of HTR-10 were simulated, and the local and global heat transfer processes were analyzed. The results show that the effects of symmetry boundary condition can be neglected when two or more water cooled pipes are included in the model. The temperature of air varies greatly near the reactor pressure vessel (RPV) and the water cooled panel. The maximum temperature occurs at the top of the RPV. With the elevation increasing, the rising rate of water temperature gradually increases. Moreover, radiation become more dominant as the RPV temperature rises.

Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10MW High Temperature gas-cooled test Reactor (HTR-10).

The R&D of HTGR high temperature helium sampling loop: From HTR-10 to HTR-PM

A High Temperature Helium Sampling Loop (HTHSL) for studying the transportation (deposition) behavior and total amount of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10MW High Temperature Gas-cooled Test Reactor (HTR-10) and High Temperature Reactor-Pebble bed Modules (HTR-PM) are researched and designed, respectively. Through the optimal design and simulation based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly so that it could be used to study fission products in the primary circuit of HTR-10. On the other hand, an improved DSD is also designed for HTR-PM based on corresponding simulations, which could be used to sample the important nuclei in the high temperature helium from SG. These schemes offer two different methods to obtain the original source term in the high temperature helium, which will provide deeper understanding for the analysis of source terms of HTGR.

Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10)

Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR). To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10) in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% forγdose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor

An optimized process for tritium-containing waste water collection of High-Temperature Gas-cooled Reactor (HTGR) was developed and experimentally verified using the 10MW High-Temperature Gas-cooled Reactor-test module (HTR-10). Compared with the previous process, an auxiliary molecular sieve bed was added in helium purification regeneration system and new operation process was proposed to collect tritium-containing waste water. In this paper, the optimized process and verification experiment were presented in detail. In commissioning experiment of the improved HTR-10, a large quantity of high-dose tritium-containing waste water was successfully collected in the water separator of helium purification regeneration system, with the specific activity being 6.1×109Bq/L. The verification experiment confirms that the optimized process is effective and reliable for the demonstration plant design of High Temperature Gas-cooled Reactor-Pebble bed module (HTR-PM) to avoid the large emission of detrimentally radioactive waste water to the environment.