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.

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.

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.

A Comprehensive Study of the 14C Source Term in the 10 MW High-Temperature Gas-Cooled Reactor

ABSTRACTWhile assessing the environmental impact of nuclear power plants, researchers have focused their attention on radiocarbon (14C) owing to its high mobility in the environment and important radiological impact on human beings. The 10 MW high-temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China that adopted helium as primary coolant and graphite spheres containing tristructural-isotropic (TRISO) coated particles as fuel elements. A series of experiments on the 14C source terms in HTR-10 was conducted: (1) measurement of the specific activity and distribution of typical nuclides in the irradiated graphite spheres from the core, (2) measurement of the activity concentration of 14C in the primary coolant, and (3) measurement of the amount of 14C discharged in the effluent from the stack. All experimental data on 14C available for HTR-10 were summarized and analyzed using theoretical calculations. A sensitivity study on the total porosity, open porosity, and percentage of closed pores that became open after irradiating the matrix graphite was performed to illustrate their effects on the activity concentration of 14C in the primary coolant and activity amount of 14C in various deduction routes.

Experimental Investigation of 14C in the Primary Coolant of the 10 MW High Temperature Gas-Cooled Reactor

ABSTRACTThe very high temperature reactor (VHTR) is a development of the high-temperature gas-cooled reactors (HTGRs) and one of the six proposed Generation IV reactor concept candidates. The 10 MW high temperature gas-cooled reactor (HTR-10) is the first pebble-bed gas-cooled test reactor in China. A sampling system for the measurement of carbon-14 (14C) was established in the helium purification system of the HTR-10 primary loop, which could sample 14C from the coolant at three locations. The results showed that activity concentration of 14C in the HTR-10 primary coolant was 1.2(1) × 102 Bq/m3 (STP). The production mechanisms, distribution characteristics, reduction routes, and release types of 14C in HTR-10 were analyzed and discussed. A theoretical model was built to calculate the amount of 14C in the core of HTR-10 and its concentration in the primary coolant. The activation reaction of 13C has been identified to be the dominant 14C source in the core, whereas in the primary coolant, it is the activation of 14N. These results can supplement important information for the source term analysis of 14C in HTR-10 and promote the study of 14C in HTGRs.

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.

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.

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).

Analysis of Seismic Soil-Structure Interaction for a Nuclear Power Plant (HTR-10)

The response of nuclear power plants (NPPs) to seismic events is affected by soil-structure interactions (SSI). In the present paper, a finite element (FE) model with transmitting boundaries is used to analyse the SSI effect on the response of NPP buildings subjected to vertically incident seismic excitation. Analysis parameters that affect the accuracy of the calculations, including the dimension of the domain and artificial boundary types, are investigated through a set of models. A numerical SSI analysis for the 10 MW High Temperature Gas Cooled Test Reactor (HTR-10) under seismic excitation was carried out using the developed model. The floor response spectra (FRS) produced by the SSI analysis are compared with a fixed-base model to investigate the SSI effect on the dynamic response of the reactor building. The results show that the FRS at foundation level are reduced and those at higher floor levels are altered significantly when taking SSI into account. The peak frequencies of the FRS are reduced due to the SSI, whereas the acceleration at high floor levels is increased at a certain frequency range. The seismic response of the primary system components, however, is reduced by the analysed SSI for the HTR-10 on the current soil site.

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.

Fabrication and Properties of Boron Carbide/Graphite Ceramic Pellets

Boron carbide (B4C)/graphite (C) ceramic pellets containing 25 wt% B4C as an ideal structural component in Chinese 10 MW high temperature gas-cooled test reactor (HTR-10) were successfully prepared by a two-stage operation method at low temperature. Initially, the samples were pressurelessly sintered at 1373 K and subsequently infiltrated with phenolic resin, pyrolysed at 1473 K, then coated with carbon by chemical vapour deposition at 1273 K. Morphology and structure of the samples were analysed by SEM and X-ray diffraction. The mechanical properties of the sintered bodies were also investigated. Superior properties were obtained due to the formation of pyrolytic carbon during the two-stage operation process. Density and compression strength of the samples were 1.94 g.cm–3 and 620 N, respectively. In addition, the rate of abrasion and moisture resistance were near zero for the samples.

The design and thermohydraulics study of the HTR-10 High Temperature Helium Experimental Loop

A High Temperature Helium Experimental Loop (HTHEL) for the purpose of studying the transportation and deposition behavior of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is studied and designed. Through the optimal design based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly and simulate the physical environment of the heat transfer tube of SG in the HTR-10 in the sense of thermohydraulics, which could be used to study the original source term in the primary circuit. The simulation of the graphite dust particle trajectories in the DSD are shown and it is elucidated that DSD could also be used to study the behavior of graphite dust in the high temperature helium in the SG of HTR, which will provide deeper understanding for the analysis of source terms of HTR-10.

The ATWS analysis of one control rod withdraw out of the HTR-10GT core in addition with bypass valve failure

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 basic 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. After a control rod in the side reflector was withdrawn out at a speed of 1cm/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, in spite of the failure of bypass valve. The residual heat would be removed out of the core by the cavity cooling system. During the accident sequence the maximum fuel temperature was 1283.3°C. It was a litter 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 will be 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 for they have the same outlet temperature of the core.