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Tatsuo Iyoku - One of the best experts on this subject based on the ideXlab platform.

  • safety demonstration tests using High Temperature Engineering test reactor
    Nuclear Engineering and Design, 2004
    Co-Authors: Shigeaki Nakagawa, Kuniyoshi Takamatsu, Yukio Tachibana, Nariaki Sakaba, Tatsuo Iyoku
    Abstract:

    Safety demonstration tests using the High Temperature Engineering test reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature gas-cooled reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005 and the second phase tests will be carried out from FY 2006.

  • Reactivity control system of the High Temperature Engineering test reactor
    Nuclear Engineering and Design, 2004
    Co-Authors: Yukio Tachibana, Tatsuo Iyoku, Hiroaki Sawahata, Toshio Nakazawa
    Abstract:

    Abstract The reactivity control system of the High Temperature Engineering test reactor (HTTR) consists of a control rod system and a reserve shutdown system. During normal operation, reactivity is controlled by the control rod system, which consists of 32 control rods (16 pairs) and 16 control rod drive mechanisms except for the case when the center control rods are removed to perform an irradiation test. In an unlikely event that the control rods fail to be inserted, reserve shutdown system is provided to insert pellets of neutron-absorbing material into the core. Alloy 800H is chosen for the metallic parts of the control rods. Because the maximum Temperature of the control rods reaches about 900 °C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the High Temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Observing the guideline, Temperature and stress analysis were conducted; it can be confirmed that the target life of the control rods of 5 years can be achieved. Various tests conducted for the control rod system and the reserve shutdown system are also described.

  • Reactor pressure vessel design of the High Temperature Engineering test reactor
    Nuclear Engineering and Design, 2004
    Co-Authors: Yukio Tachibana, Shigeaki Nakagawa, Tatsuo Iyoku
    Abstract:

    The reactor pressure vessel (RPV) of the HTTR is 5.5 m (inside diameter), 13.2 m (inside height), and 122 mm (shell thickness). The RPV contains core components, reactor internals, reactivity control system, etc. 2 1/4Cr–1Mo steel is chosen as the material for RPV. The Temperature reaches about 400 °C at normal operation. The fluence of the RPV is estimated to be less than 1 × 1017 n/cm2 (E > 1 MeV) and so irradiation embrittlement is negligible, but temper embrittlement is not negligible. For the purpose of reducing embrittlement, content of some elements must be limited in the 2 1/4Cr–1Mo steel for the RPV; embrittlement parameters, J  -factor and X¯ are used. In this paper, design and structure of the RPV are reviewed first. Fabrication procedure of the RPV and its special feature are described. Material data on the 2 1/4Cr–1Mo steel manufactured for the RPV, especially the embrittlement parameters, J  -factor and X¯, and nil-ductility transition Temperatures, TNDT, by drop weight tests, are shown. In-service inspection and results of R&Ds are also described.

  • Design and Fabrication of Reactor Pressure Vessel for High Temperature Engineering Test Reactor (HTTR)
    Elevated Temperature Design and Analysis Nonlinear Analysis and Plastic Components, 2004
    Co-Authors: Yukio Tachibana, Shigeaki Nakagawa, Tatsuo Iyoku
    Abstract:

    The reactor pressure vessel (RPV) of the HTTR is 5.5 m in inside diameter, 13.2 m in inside height, and 122 mm and 160 mm in wall thickness of the body and the top head dome, respectively. Because the reactor inlet Temperature of the HTTR is Higher than that of LWRs, 2 1/4Cr-1Mo steel is chosen for the RPV material. Fluence of the RPV is estimated to be less than 1×1017 n/cm2 (E>1 MeV), and so irradiation embrittlement is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the RPV using embrittlement parameters, J-factor and X . In this paper design, fabrication procedure, and in-service inspection technique of the RPV for the HTTR are described.Copyright © 2004 by ASME

  • plan for first phase of safety demonstration tests of the High Temperature Engineering test reactor httr
    Nuclear Engineering and Design, 2003
    Co-Authors: Yukio Tachibana, Takeshi Takeda, Shigeaki Nakagawa, Kazuhiro Sawa, Kuniyoshi Takamatsu, Akio Saikusa, Takayuki Furusawa, Tatsuo Iyoku
    Abstract:

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).

Shusaku Shiozawa - One of the best experts on this subject based on the ideXlab platform.

  • Fabrication of the First-Loading Fuel of the High Temperature Engineering Test Reactor
    Journal of Nuclear Science and Technology, 1999
    Co-Authors: Kazuhiro Sawa, Shusaku Shiozawa, Tsutomu Tobita, Haruyoshi Mogi, Shigeharu Yoshimuta, Shuuichi Suzuki, Kouzaburou Deushi
    Abstract:

    The High Temperature Engineering Test Reactor (HTTR). which is the first High Temperature gas-cooled reactor (HTGR) in Japan, attained its first criticality in November 1998. The fabrication of the...

  • Integrity assessment of the High Temperature Engineering test reactor (HTTR) control rod at very High Temperatures
    Nuclear Engineering and Design, 1997
    Co-Authors: Yukio Tachibana, Shusaku Shiozawa, Juichi Fukakura, F. Matsumoto, T. Araki
    Abstract:

    The High Temperature Engineering test reactor (HTTR) is the first High Temperature gas-cooled reactor (HTGR) in Japan with a reactor outlet coolant Temperature of 950°C at High Temperature test operation. The HTTR contains 16 pairs of control rods for which Alloy 800H is chosen of the metallic parts. Because the maximum Temperature of the control rods reaches about 900°C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the High Temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Under the guideline, Temperature and stress analysis was conducted, and it is confirmed that the target life of the control rods of 5 years can be achieved.

  • Present status of High-Temperature Engineering Test Reactor (HTTR) program
    Transactions of the American Nuclear Society, 1994
    Co-Authors: Toshiyuki Tanaka, Shusaku Shiozawa, Osamu Baba, Minoru Okubo, Toshiaki Tobioka
    Abstract:

    The 30MWt HTTR is a High-Temperature gas-cooled reactor (HTGR), with a maximum helium coolant Temperature of 950 degreesC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel(RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories aiming at their installation starting in the middle of 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERT) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant Temperature. This paper describes major features of HTTR, present status of its construction and research and test program using HTTR.

  • Seismic Response of the High-Temperature Engineering Test Reactor Core Bottom Structure
    Nuclear Technology, 1992
    Co-Authors: Tatsuo Iyoku, Shusaku Shiozawa, Masatoshi Futakawa, Yoshiyuki Inagaki, Toshiyo Miki
    Abstract:

    This paper discusses the High-Temperature Engineering Test Reactor (HTTR) a 30-MW (thermal) helium gas-cooled rector that uses a prismatic block. The core bottom structure (CBS) of the HTTR consists of an arrangement of graphite components, and it supports the core elements within the reactor vessel. vibration tests are performed with two scale models to clarify the seismic response of the CBS. The vibration characteristics of the CBS are clarified quantitatively, and the structural integrity of the graphite components is confirmed.

  • Seismic Response of the High-Temperature Engineering Test Reactor Core Bottom Structure
    Nuclear Technology, 1992
    Co-Authors: Tatsuo Iyoku, Shusaku Shiozawa, Masatoshi Futakawa, Yoshiyuki Inagaki, Toshiyo Miki
    Abstract:

    This paper discusses the High-Temperature Engineering Test Reactor (HTTR) a 30-MW (thermal) helium gas-cooled rector that uses a prismatic block. The core bottom structure (CBS) of the HTTR consists of an arrangement of graphite components, and it supports the core elements within the reactor vessel. vibration tests are performed with two scale models to clarify the seismic response of the CBS. The vibration characteristics of the CBS are clarified quantitatively, and the structural integrity of the graphite components is confirmed.

Kazuhiro Sawa - One of the best experts on this subject based on the ideXlab platform.

  • Burn-Up Dependency of Control Rod Position at Zero-Power Criticality in the High-Temperature Engineering Test Reactor
    Journal of Nuclear Engineering and Radiation Science, 2016
    Co-Authors: Yuki Honda, Nozomu Fujimoto, Shoji Takada, Hiroaki Sawahata, Kazuhiro Sawa
    Abstract:

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

  • Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor
    Journal of Nuclear Engineering and Radiation Science, 2016
    Co-Authors: Yuki Honda, Nozomu Fujimoto, Shoji Takada, Hiroaki Sawahata, Kazuhiro Sawa
    Abstract:

    The High-Temperature Engineering test reactor (HTTR) is a block-type High-Temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core Temperature of the HTTR is too High. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross-section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

  • Investigation on Iodine Release Behavior During the Operation of High Temperature Engineering Test Reactor (HTTR)
    Volume 2: Plant Systems Construction Structures and Components; Next Generation Reactors and Advanced Reactors, 2013
    Co-Authors: Shohei Ueta, Nariaki Sakaba, Hiroyuki Inoi, Yoshitaka Mizutani, Hirofumi Ohashi, Jin Iwatsuki, Kazuhiro Sawa
    Abstract:

    Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.Copyright © 2013 by ASME

  • fuel and fission gas behavior during rise to power test of the High Temperature Engineering test reactor httr
    Journal of Nuclear Science and Technology, 2003
    Co-Authors: Shohei Ueta, Junya Sumita, Koichi Emori, Masashi Takahashi, Kazuhiro Sawa
    Abstract:

    The rise-to-power tests of the High Temperature Engineering Test Reactor (HTTR) have been carried out successfully by the Japan Atomic Energy Research Institute (JAERI). For the safe operation of HTTR, the continuous and reliable measurement of the coolant activity is required to evaluate the fuel performance during normal operating conditions. In order to measure the primary coolant radioactivity (PCR), the PCR instrumentation of the safety protection system, the fuel failure detection (FFD) system and the primary coolant sampling system have been installed in the primary circuit. The PCR was less than 103 MBq/m3, and measured Kr and Xe isotopes were less than 0.1 MBq/m3 during the rise-to-power tests. The measured fractional releases are constant at 2x10-9 up to 60% of the reactor power, and then increase to 7x10-9 at full power operation. The prediction shows good agreement with the measured value. These results showed that the release mechanism varied from recoil to diffusion of the generated fission ...

  • plan for first phase of safety demonstration tests of the High Temperature Engineering test reactor httr
    Nuclear Engineering and Design, 2003
    Co-Authors: Yukio Tachibana, Takeshi Takeda, Shigeaki Nakagawa, Kazuhiro Sawa, Kuniyoshi Takamatsu, Akio Saikusa, Takayuki Furusawa, Tatsuo Iyoku
    Abstract:

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).

Shoji Takada - One of the best experts on this subject based on the ideXlab platform.

  • Promising Neutron Irradiation Applications at the High Temperature Engineering Test Reactor
    Journal of Nuclear Engineering and Radiation Science, 2020
    Co-Authors: Yuki Honda, Nozomu Fujimoto, Shimpei Hamamoto, Toshiaki Ishii, Shoji Takada, Etsuo Ishitsuka
    Abstract:

    Abstract The High Temperature gas-cooled reactor (HTGR) has advantages for irradiation applications such as large space available for irradiation at reflector region and High thermal neutron spectrum with the graphite moderator. High Temperature Engineering test reactor (HTTR), a prismatic type of the HTGR, has been constructed to establish and upgrade the basic technologies for the HTGRs. Many irradiation regions are reserved in the HTTR to be served as a potential tool for an irradiation test reactor in order to promote innovative basic researches such as materials, fusion reactor technology, and radiation chemistry. This study shows the overview of some possible irradiation applications at the HTTRs including neutron transmutation doping silicon (NTD-Si) and Iodine-125 (125I) productions. The HTTR has possibility to produce about 40 tons of doped Si-particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8 × 105 GBq/yr of 125I isotope, comparing to 3.0 × 103 GBq of total 125I supplied in Japan in 2016.

  • Loss of core cooling test with one cooling line inactive in Vessel Cooling System (VCS) of High Temperature Engineering Test Reactor (HTTR)
    Journal of Nuclear Engineering and Radiation Science, 2017
    Co-Authors: Yusuke Fujiwara, Daisuke Tochio, Masanori Shinohara, Takahiro Nemoto, Masato Ono, Shoji Takada
    Abstract:

    In the High-Temperature Engineering test reactor (HTTR), the vessel cooling system (VCS) which is arranged around the reactor pressure vessel (RPV) removes residual heat and decay heat from the reactor core when the forced core cooling is lost. The test of loss of forced cooling (LOFC) when one of two cooling lines in VCS lost its cooling function was carried out to simulate the partial loss of cooling function from the surface of RPV using the HTTR at the reactor thermal power of 9 MW, under the condition that the reactor power control system and the reactor inlet coolant Temperature control system were isolated, and three helium gas circulators (HGCs) in the primary cooling system (PCS) were stopped. The test results showed that the reactor power immediately decreased to almost zero, which is caused by negative feedback effect of reactivity, and became stable as soon as HGCs were stopped. On the other hand, the Temperature changes of permanent reflector block, RPV, and the biological shielding concrete were quite slow during the test. The Temperature decrease of RPV was several degrees during the test. The numerical result showed a good agreement with the test result of Temperature rise of biological shielding concrete around 1 °C by the numerical method that uses a calibrated thermal resistance by using the measured Temperatures of RPV and the air outside of biological shielding concrete. The Temperature increase of water cooling tube panel of VCS was calculated to be about 15 °C which is sufficiently small in the view point of property protection. It was confirmed that the sufficient cooling capacity of VCS can be maintained even in case that one of two water cooling lines of VCS loses its function.

  • improvement of neutron startup source handling work by developing new transportation container for High Temperature Engineering test reactor httr
    Journal of Nuclear Science and Technology, 2017
    Co-Authors: Yosuke Shimazaki, Hiroaki Sawahata, Masanori Shinohara, Yoshinori Yanagida, Taiki Kawamoto, Shoji Takada
    Abstract:

    ABSTRACTThe High-Temperature Engineering test reactor (HTTR) has three neutron startup sources (NSs) in the reactor core, each of which consists of 252Cf with 3.7 GBq, installed in NS holder and subsequently in a control rod guide block (CR block). The NSs are exchanged at the interval of approximately seven years. The NS holders are transported from the dealer's hot cell to the HTTR using a transportation container. The loading work of NS holders to the CR blocks is subsequently carried out in the fuel handling machine maintenance pit of HTTR.Technical issues, which are the reduction and prevention of radiation exposure of workers and the exclusion of falling of NS holder, were extracted from the experiences in the past two exchange works of NSs to develop a safety handling procedure. Then, a new transportation container special to the NSs of HTTR was developed to solve the technical issues while keeping the cost as low as that for overhaul of conventional container.As a result, the NS handling work usin...

  • Burn-Up Dependency of Control Rod Position at Zero-Power Criticality in the High-Temperature Engineering Test Reactor
    Journal of Nuclear Engineering and Radiation Science, 2016
    Co-Authors: Yuki Honda, Nozomu Fujimoto, Shoji Takada, Hiroaki Sawahata, Kazuhiro Sawa
    Abstract:

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

  • Study on Sensitivity of Control Rod Cell Model in Reflector Region of High-Temperature Engineering Test Reactor
    Journal of Nuclear Engineering and Radiation Science, 2016
    Co-Authors: Yuki Honda, Nozomu Fujimoto, Shoji Takada, Hiroaki Sawahata, Kazuhiro Sawa
    Abstract:

    The High-Temperature Engineering test reactor (HTTR) is a block-type High-Temperature gas-cooled reactor (HTGR). There are 32 control rods (16 pairs) in the HTTR. Six of the pairs of control rods are located in a core region and the remainder are located in a reflector region surrounding the core. Inserting all control rods simultaneously at the reactor scram in a full-power operation presents difficulty in maintaining the integrity of the metallic sleeve of the control rod because the core Temperature of the HTTR is too High. Therefore, a two-step control rod insertion method is adopted for the reactor scram. The calculated control rod worth at the first step showed a larger underestimation than the measured value in the second step, although the calculated results of the excess reactivity tests showed good agreement with the measured result in the criticality tests of the HTTR. It is concluded that a cell model for the control rod guide block with the control rod in the reflector region is not suitable. In addition, in the core calculation, the macroscopic cross section of a homogenized region of the control rod guide block with the control rod is used. Therefore, it would be one of the reasons that the neutron flux distribution around the control rod in control rod guide block in the reflector region cannot be simulated accurately by the conventional cell model. In the conventional cell model, the control rod guide block is surrounded by the fuel blocks only, although the control rods in the reflector region are surrounded by both the fuel blocks and the reflector blocks. The difference of the neutron flux distribution causes the large difference of a homogenized macroscopic cross-section set of the control rod guide block with the control rod. Therefore, in this paper, the cell model is revised for the control rod guide block with the control rod in the reflector region to account for the actual configuration around the control rod guide block in the reflector region. The calculated control rod worth at the first step using the improved cell model shows better results than the previous one.

Shigeaki Nakagawa - One of the best experts on this subject based on the ideXlab platform.

  • Li-rod structure in High-Temperature gas-cooled reactor as a tritium production device for fusion reactors
    Fusion Engineering and Design, 2019
    Co-Authors: Hideaki Matsuura, Shigeaki Nakagawa, Etsuo Ishitsuka, Minoru Goto, Ryo Okamoto, Yuki Koga, Takuro Suganuma, Kazunari Katayama, Teppei Otsuka, Kenji Tobita
    Abstract:

    Abstract Production of tritium using a High-Temperature gas-cooled reactor (HTGR) has been studied for a prior Engineering test with tritium handling and for the startup operation of a demonstration fusion reactor. For this purpose, the hydrogen absorption speed of Zr in a Li-loading rod for the reactor operation is experimentally measured, and an analysis model is presented to evaluate the tritium outflow from the Li rod in a High-Temperature Engineering test reactor (HTTR). On the basis of the presented model, the structure of the Li-loading rod for the demonstration test using the HTTR is proposed.

  • Actual Operation and Control of High-Temperature Engineering Test Reactor
    Nuclear Reactor Kinetics and Plant Control, 2012
    Co-Authors: Shigeaki Nakagawa
    Abstract:

    The High-Temperature Engineering test reactor (HTTR) (Fig. 10.1) is the first High-Temperature gas-cooled reactor in Japan. It first attained criticality in November 1998, and the reactor outlet Temperature reached 850°C in December 2001, and 950°C in June 2004, both world firsts.

  • Numerical simulation of three-dimensional thermal-hydraulic behavior for HTTR (High Temperature Engineering Test Reactor)
    Nuclear Engineering and Design, 2011
    Co-Authors: Daisuke Tochio, Shigeaki Nakagawa
    Abstract:

    Abstract Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are now in progress in order to verify the inherent safety features and to improve safety designing and analysis technologies for future HTGR (High Temperature gas-cooled reactor). Coolant flow reduction test is one of the safety demonstration tests for the purpose of demonstration of inherent HTGR safety features in the case that coolant flow is reduced by tripping of helium gas circulators. If reactor core element Temperature and core internal structure Temperature during abnormal events are estimated by numerical simulation with High-accuracy, developed numerical simulation method can be applied to future HTGR designing efficiently. In the present research, three-dimensional in- and ex-vessel thermal-hydraulic calculations for the HTTR are performed with a commercially available thermal-hydraulic analysis code “STAR-CD ® ” with finite volume method. The calculations are performed for normal operation and coolant flow reduction tests of the HTTR. Then calculated Temperatures are compared with measured ones obtained in normal operation and coolant flow reduction test. As the result, calculated Temperatures are good agreement with measured ones in normal operation and coolant flow reduction test.

  • Core Dynamics Analysis for Reactivity Insertion and Loss of Coolant Flow Tests Using the High Temperature Engineering Test Reactor
    Journal of Power and Energy Systems, 2008
    Co-Authors: Kuniyoshi Takamatsu, Shigeaki Nakagawa, Tetsuaki Takeda
    Abstract:

    Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named-ACCORD-, was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We have modified this code to use a model with four parallel channels and twenty Temperature coefficients. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results. Moreover, the effect of the model is formulated quantitatively with our proposed equation. Finally, the pre-analytical result of the loss of coolant flow test by tripping all gas circulators is also discussed.

  • safety demonstration tests using High Temperature Engineering test reactor
    Nuclear Engineering and Design, 2004
    Co-Authors: Shigeaki Nakagawa, Kuniyoshi Takamatsu, Yukio Tachibana, Nariaki Sakaba, Tatsuo Iyoku
    Abstract:

    Safety demonstration tests using the High Temperature Engineering test reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature gas-cooled reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005 and the second phase tests will be carried out from FY 2006.