Temperature Gas-cooled Reactor-Test Module Research Articles

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.

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.

Decomposed Functional Behavior of Helium Purification Support System for Experimental Power Reactor (RDE) Type Using Sequence Diagram

Instrumentation and Control (I&C) system in Nuclear Power Plant (NPP) has a key role to maintaining the safety and control the integrated system to operate the NPP. The aim of helium purification instrumentation system in High Temperature Gas-cooled Reactor-test Module (HTR-10) reactor is to reduce chemical impurity levels in the primary helium cooling system and to eliminate the fission products of radionuclide gas. Indonesia conduct a research about Experimental Power Reactor (RDE) with HTR-10 type of reactor, then the helium purification system has an important role since this reactor will use helium in the primary cooling system. Since this HTR-10 type of reactor is had been develop by Tsinghua University in China since 1995, then for development of RDE in Indonesia should refer to the initial design of HTR-10 from China, and continuing to improve the design by decomposed the system function so that RDE design will more safety and reliable. After decomposed the new system of helium purification, the sequence diagram of helium purification will developed to review the new system function of helium purification instrumentation. Sequence diagram is system engineering methodology which can help the design engineer or system engineer to understand the system behavior by mention in to important requirement of the system. By this sequence diagram, it hope that the design engineer will clearly understand the functions and translate in to design synthesis and architectural design in the next step of development phase of helium purification I&C system.

Reverse Engineering Program Using MBSE to Support Development of I&C System Experimental Power Reactor from PLC to FPGA.

Indonesian Experimental Power Reactor (RDE) basic design is designed by refer to the High Temperature Gas-cooled reactor-test Module (HTR-10) type which developed by Tsinghua University in China since 1995 and receiving the first critically on December 2000. Currently, the market for a control system in nuclear power plant (NPP) industry is using micro-controller and Programmable Logic Controller (PLC). Since the computer-based technologies are vulnerable to cyber-attack possibility, software common cause failure (CCF) and system complexity, however, development of RDE designs should be thought of by the latest technologies in accordance with the development of Instrumentation and Control (I&C) system which play an important role in maintaining the safety aspect for NPP. This study deals with the reverse engineering program for I&C with the PLC-based system, to obtain the design specification from the previous design in order to improve its reliability by considering systems hardware using Field-programmable Gate Array (FPGA) as an alternative platform. Before developing reverse engineering should analyze why FPGA become an alternative system to change the PLC-based system. Reverse engineering process will cover by model-based system engineering (MBSE), which is the formal modeling application to support systems requirements, design, analysis, verification and validation (V&V) activities. The process starting with the conceptual design, requirement analysis, continuing throughout development and later life cycle phases which is recommended using V model. It is hoped that by doing this reverse engineering program, there is an analysis function related I&C system so that the more reliable design with the use of FPGA-based platform can be made. Reverse engineering can also help to track back the original development with I&C platform from the licensing requirement so that to develop re-engineering program for the new FPGA-based I&C system can meet the design criteria and accepted by the regulatory commission. The requirement, selection of lifecycle model, and verification methodology will be will be explained clearly in order to become a reference in the development plan of RDE in Indonesia.

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.

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.

Post-test simulation of the HTR-10 reactivity insertion without scram

The 10MW high temperature gas-cooled reactor-test module (HTR-10) is the first high temperature gas-cooled reactor (HTGR) in China. After the commissioning phase, several safety demonstration tests, focusing on simulating the anticipated transient without scram (ATWS) scenarios, were successfully performed on this reactor. Among these tests, a reactivity insertion ATWS test was conducted in November, 2003 by withdrawing a single control rod with no activation of the reactor shutdown system. This test was carried out at 30% rated power, i.e., 3MW power level, with about 1‰ Δk/k positive reactivity introduced into the core.In this study, the HTR-10 dynamic response during the reactivity insertion ATWS test is simulated and analyzed with the help of the THERMIX code. The inherent safety features of the HTR-10 are revealed and verified by both the test and the simulation results. Important physical phenomena, e.g., automatic shutdown, reactor recriticality, power oscillation, helium natural circulation and temperature redistribution, are investigated. Based on the actual test conditions, the simulation reproduces the reactor transients very well. For the reactor power and the temperatures of its internals, the code results are compared with their corresponding measured values. Good agreement between the calculation results and the test results proves the THERMIX simulation capability for the HTR-10 transient behavior during the reactivity insertion ATWS test. As the most important safety-related parameter, the maximum fuel temperature during the test process is always much lower than its safety limit of 1620°C.

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.

Temperature Response of the HTR-10 during the Power Ascension Test

The 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) is the first High Temperature Gas-Cooled Reactor in China. With the objective of raising the reactor power from 30% to 100% rated power, the power ascension test was planned and performed in January 2003. The test results verified the practicability and validity of the HTR-10 power regulation methods. In this study, the power ascension process is preliminarily simulated using the THERMIX code. The code satisfactorily reproduces the reactor transient parameters, including the reactor power, the primary helium pressure, and the primary helium outlet temperature. Reactor internals temperatures are also calculated and compared with the test values recorded by a number of thermocouples. THERMIX correctly simulates the temperature variation tendency for different measuring points, with good to fair agreement between the calculated temperatures and the measured ones. Based on the comparison results, the THERMIX simulation capability for the HTR-10 dynamic characteristics during the power ascension process can be demonstrated. With respect to the reactor safety features, it is of utmost importance that the maximum fuel center temperature during the test process is always much lower than the fuel temperature limit of 1620°C.

Benchmark Calculation for the Steady-State Temperature Distribution of the HTR-10 under Full-Power Operation

Within the framework of a Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor Performance (CRP-5) initiated by the International Atomic Energy Agency (IAEA), the calculation of steady-state temperature distribution of the 10MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) under its initial full power experimental operation has been defined as one of the benchmark problems. This paper gives the investigation results obtained by different countries who participate in solving this benchmark problem. The validation works of the THERMIX code used by the Institute of Nuclear and New Energy Technology (INET) are also presented. For the benchmark items defined in this CRP, various calculation results correspond well with each other and basically agree the experimental results. Discrepancies existing among various code results are preliminarily attributed to different methods, models, material properties, and so on used in the computations. Temperatures calculated by THERMIX for the measuring points in the reactor internals agree well with the experimental values. The maximum fuel center temperatures calculated by the participants are much lower than the limited value of 1,230°C. According to the comparison results of code-to-code as well as code-toexperiment, THERMIX is considered to reproduce relatively satisfactory results for the CRP-5 benchmark problem.

Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10

Safety demonstration tests were conducted on the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) to verify the inherent safety characteristics of modular High Temperature Gas-cooled Reactors (HTGRs) as well as to obtain the transient data of reactor core and primary cooling system for validation of HTGR safety analysis models and codes. As one of these safety demonstration tests, a simulated anticipated transient without scram (ATWS) test called loss of forced cooling by tripping the helium circulator without reactor scram was carried out at 3 MW power level on October 15, 2003. This paper simulates and analyzes the power transient and the thermal response of the reactor during the test by using the THERMIX code. The analytical results are compared with the test data for validation of the code. Owing to the negative temperature coefficient of reactivity, the reactor undergoes a self-shut down after the stop of the helium circulator; the subsequent phenomena such as the recriticality and power oscillations are also studied. During the test a natural circulation loop of helium is established in the core and the other coolant channels and its consequent thermal response such as the temperature redistribution is investigated. In addition, temperatures of the measuring points in the reactor internals are calculated and compared with the measured values. Satisfactory agreements obtained from the comparison demonstrate the basic applicability and reasonability of the THERMIX code for simulating and analyzing the helium circulator trip ATWS test. With respect to the safety features of the HTR-10, it is of most importance that the maximum fuel center temperature during the test is always lower than 1600 °C which is the limited value for the HTGR.

Benchmark Calculation for the Steady-State Temperature Distribution of the HTR-10 under Full-Power Operation

Within the framework of a Coordinated Research Project on Evaluation of High Temperature Gas-Cooled Reactor Performance (CRP-5) initiated by the International Atomic Energy Agency (IAEA), the calculation of steady-state temperature distribution of the 10MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10) under its initial full power experimental operation has been defined as one of the benchmark problems. This paper gives the investigation results obtained by different countries who participate in solving this benchmark problem. The validation works of the THERMIX code used by the Institute of Nuclear and New Energy Technology (INET) are also presented. For the benchmark items defined in this CRP, various calculation results correspond well with each other and basically agree the experimental results. Discrepancies existing among various code results are preliminarily attributed to different methods, models, material properties, and so on used in the computations. Temperatures calculated by THERMIX for the measuring points in the reactor internals agree well with the experimental values. The maximum fuel center temperatures calculated by the participants are much lower than the limited value of 1,230°C. According to the comparison results of code-to-code as well as code-toexperiment, THERMIX is considered to reproduce relatively satisfactory results for the CRP-5 benchmark problem.

06/01214 The feasibility study on perfect burning reactor system: Nakae, N. Progress in Nuclear Energy, 2005, 47, (1–4), 139–146

In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced.

Transient tests on blower trip and rod removal at the HTR-10

Safety demonstration tests on the 10 MW high temperature gas-cooled reactor test module (HTR-10) were conducted to verify the inherent safety features of MHTGRs and to obtain the core and primary cooling system transient data for validation of safety analysis codes. Two simulated anticipated transients without scram (ATWS) tests, lose of forced cooling by trip of the helium blower and reactivity insertion via control rod withdrawal were performed. This paper describes the tests with detailed test method, condition and results. Calculated results show that the strongly negative temperature coefficient causes reactor power to closely follow heat removal levels. Maximum fuel temperature changes are limited by the large core heat capacity to below 1230 °C during two tests. The test of tripping the helium circulator ATWS test was conducted on October 15, 2003. Although none of 10 control rods was moved, the reactor power immediately decreased due to the negative temperature coefficient. After about 50 min, the reactor became criticality again. Finally, the reactor power went to a stable level with about 200 kW. The test of reactivity insertion ATWS test was conducted two times. Following the control rod withdrawal, the reactor power increased rapidly, the maximum power level reached to 5037 and 7230 kW from the initial power of 3000 kW in accordance with reactivity insertion of $ 0.136 and 0.689, respectively. After the reactivity introduced was compensated by means of the strong negative reactivity feedback effect, the reactor went to subcritical and the power decreased.

High temperature reactor development in China

Since the late 1970'-s the research and development program on the high temperature gas-cooled reactor (HTR) has been carried out in China. The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) reached first criticality in 2000 and was put into full power operation in 2003. Six safety demonstration tests were done on the HTR-10. The project of the HTR-10 with a gas turbine cycle is underway. The project of the HTR demonstration plant with a power of around 150 MWe (HTR-PM) is planned. In this paper the HTR development in China is briefly described.

Probability assessment of graphite brick in the HTR-10

A probability finite element assessment program was developed to evaluate the security of graphite components in the HTR-10 (10 MW high temperature gas-cooled reactor-test module), based on the MARC non-linear finite element code and the strength uncertainty of the graphite material. Using user-defined subroutines (UDS), the irradiation thermal analysis subroutine, irradiation static analysis subroutine and probability assessment subroutine are embedded into the MARC program. The recompiled MARC program take into account irradiation-induced changes in graphite components such as the thermal conductivity coefficient, the thermal expansion coefficient, the creep coefficient, the elastic modulus, and the strength. The failure probabilities of the graphite components in the HTR-10, either under normal operating conditions or cold shutdown conditions, were evaluated. Additional analyses were done with the irradiation deformation increasing 20% and the creep coefficient decreasing 20%, to see the influence of irradiation deformation and the creep effect on the failure probability. The study showed that the probability finite element assessment method is an effective tool to assess the probability of structure failure.

03/02255 Overview of the 10 MW high temperature gas cooled reactor-test module project: Xu, Y. and Zuo, K. Nuclear Engineering and Design, 2002. 218, (1–3). 13–23

03/02255 Overview of the 10 MW high temperature gas cooled reactor-test module project: Xu, Y. and Zuo, K. Nuclear Engineering and Design, 2002. 218, (1–3). 13–23

Prospects for economic fusion electricity : Cook, I. et al. Fusion Engineering and Design, 2002, 63–64, 25–33

Since the late 1970'-s the research and development program on the high temperature gas-cooled reactor (HTR) has been carried out in China. The 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) reached first criticality in 2000 and was put into full power operation in 2003. Six safety demonstration tests were done on the HTR-10. The project of the HTR-10 with a gas turbine cycle is underway. The project of the HTR demonstration plant with a power of around 150 MWe (HTR-PM) is planned. In this paper the HTR development in China is briefly described.

Design and full scale test of the fuel handling system

In the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) fuel elements move through the core driven by gravity. To reach their design burn-up the fuel elements are re-shuttled five times. This transportation outside the core is mainly achieved pneumatically. Although, adopting the international experience at design and operation of similar systems some key components were improved so that the fuel handling system (FHS) becomes simpler and more reliable. The improved components were tested in full-scale testing facilities. The debugging test and the first loading operation for the FHS indicate that the FHS meets the demands of the HTR-10. In this paper, the functions, design parameters, technological processes, main components and design characteristics of the FHS are described in detail. The flow schemes, design parameters of the full-scale testing facilities and the experimental results are briefly introduced.

Design and experiment of hot gas duct for the HTR-10

A horizontal coaxial double-tube hot gas duct is a key component connecting the reactor pressure vessel and the steam generator pressure vessel for the 10 MW High Temperature Gas-cooled Reactor—Test Module. Hot helium gas from the core outlet flows into the steam generator through the liner tube, while helium gas after being cooled returns to the core through a passage formed between the inner tube and the duct pressure vessel. Thermal insulation material is packed into the space between the liner tube and the inner tube to resist heat transfer from the hot helium to the cold helium. The thermal compensation structure is designed in order to avoid large thermal stress because of different thermal expansions of the duct parts under various conditions. According to the design principal of the hot gas duct, the detailed structure design and strength evaluation for it has been done. A full-scale duct test section was then made according to the design parameters, and its thermal performance experiment was carried out in a helium test loop. With helium gas at pressure of about 3.0 MPa and a temperature over 900 °C, the continuous operation time for the duct test section lasted 98 h. At a helium gas temperature over 700 °C, the cumulative operation time for the duct test section reached 350 h. The duct test section also experienced 20 pressure cycles in the pressure range of 0.1–3.4 MPa, 18 temperature cycles in the temperature range of 100–950 °C. Thermal test results show an effective thermal conductivity of the hot gas duct thermal insulation is 0.47 W m −1 °C −1 under normal operation condition. In addition, a hot gas duct depressurization test was carried out; the test result showed that the pressure variation occurred on the liner tube was not more than 0.2 MPa for an assumed maximum gas release rate.