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Takada, Shoji; Ngarayana, I. W.*; Nakatsuru, Yukihiro*; Terada, Atsuhiko; Murakami, Kenta*; Sawa, Kazuhiro*
Mechanical Engineering Journal (Internet), 7(3), p.19-00536_1 - 19-00536_12, 2020/06
In this study reasonable 2D model was established by using FLUENT for start-up of analysis and evaluation of heat transfer flow characteristics in 1/6 scale model of VCS for HTTR. By setting up pressure vessel temperature around 200
C about relatively high ratio of heat transfer via natural convection in total heat removal around 20-30%, which is useful for code to experiment benchmark in the aspect to confirm accuracy to predict temperature distribution of components which is heated up by natural convection flow. The numerical results of upper head of pressure vessel by the 
-
-SST intermittency transition model, which can adequately reproduce the separation, re-adhesion and transition, reproduced the test results including temperature distribution well in contrast to those by the -
model in both cases that helium gas is evacuated or filled in the pressure vessel. It was emerged that any local hot spot did not appear on the top of upper head of pressure vessel where natural convection flow of air is separated in both cases. In addition, the plume of high temperature helium gas generated by the heating of heater was well mixed in the upper head and uniformly heated the inner surface of upper head without generating hot spots.
Ishitsuka, Etsuo; Matsunaka, Kazuaki*; Ishida, Hiroki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Kondo, Atsushi*; et al.
JAEA-Technology 2019-008, 12 Pages, 2019/07
JAEA-Technology-2019-008.pdf:2.37MB
As a summer holiday practical training 2018, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out. As a result, it is become clear that the continuous operations for about 30 years at 2 MW, about 25 years at 3 MW, about 18 years at 4 MW, about 15 years at 5 MW are possible. As an image of thermal design, the image of the nuclear battery consisting a cooling system with natural convection and a power generation system with no moving equipment is proposed. Further feasibility study to confirm the feasibility of nuclear battery will be carried out in training of next fiscal year.
Takamatsu, Kuniyoshi
Annals of Nuclear Energy, 106, p.71 - 83, 2017/08
The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.
Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju
JAEA-Data/Code 2017-002, 74 Pages, 2017/03
JAEA-Data-Code-2017-002.pdf:2.36MB
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Inaba, Yoshitomo; Nishihara, Tetsuo
Annals of Nuclear Energy, 101, p.383 - 389, 2017/03
Times Cited Count：8 Percentile：57.53(Nuclear Science & Technology)In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Tochio, Daisuke; Honda, Yuki; Sato, Hiroyuki; Sekita, Kenji; Homma, Fumitaka; Sawahata, Hiroaki; Takada, Shoji; Nakagawa, Shigeaki
Journal of Nuclear Science and Technology, 54(1), p.13 - 21, 2017/01
Times Cited Count：2 Percentile：17.85(Nuclear Science & Technology)GTHTR300C is designed and developed in JAEA. The reactor system is required to continue a stable and safety operation as well as a stable power supply in the case that thermal-load is fluctuated by the occurrence of abnormal event in the heat utilization system. Then, it is necessary to demonstrate that the thermal-load fluctuation should be absorbed by the reactor system so as to continue the stable and safety operation could be continued. The thermal-load fluctuation absorption tests without nuclear heating were planned and conducted in JAEA to clarify the absorption characteristic of thermal-load fluctuation mainly by the reactor and by the IHX. As the result it was revealed that the reactor has the larger absorption capacity of thermal-load fluctuation than expected one, and the IHX can be contributed to the absorption of the thermal-load fluctuation generated in the heat utilization system in the reactor system. It was confirmed from there result that the reactor and the IHX has effective absorption capacity of the thermal-load fluctuation generated in the heat utilization system. Moreover it was confirmed that the safety estimation code based on RELAP5/MOD3 can represents the thermal-load fluctuation absorption behavior conservatively.
Tochio, Daisuke; Fujimoto, Nozomu
Journal of Nuclear Science and Technology, 53(3), p.425 - 431, 2016/03
Times Cited Count：1 Percentile：9.76(Nuclear Science & Technology)The future HTGR is now designed in JAEA. The reactor has many merging points of helium gas with different temperature. It is needed to clear the mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the HTTR due to lack of mixing of helium gas in the primary cooling system. Now the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal-hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the mixing behavior of helium gas. As the result, it was confirmed that the mixing behavior of helium gas in the primary cooling system is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.
Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05
For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.
Nishihara, Tetsuo; Inagaki, Yoshiyuki
Nuclear Technology, 153(1), p.100 - 106, 2006/01
Times Cited Count：9 Percentile：52.57(Nuclear Science & Technology)Japan Atomic Energy Research Institute (JAERI) has performed the research and development of hydrogen production using the high temperature engineering test reactor (HTTR). One of the key issues for the HTTR hydrogen production system is the development of control technology for stable operation. A thermal load absorber concept using a steam generator installed downstream of a reformer is proposed to mitigate a variation of helium temperature. Thermal hydraulic analyses for the start up operation and the suspension of feed gas supply to the reformer are carried out. These results show that a large variation of the reformer outlet helium temperature takes place due to a change of the feed gas flow rate. However the steam generator can mitigate the variation of helium temperature. It is clarified that the HTTR can continue normal operation independently of the feed gas flow rate.
Sumita, Junya; Shibata, Taiju; Nakagawa, Shigeaki; Hanawa, Satoshi; Iyoku, Tatsuo; Ishihara, Masahiro
Transactions of 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT-18), p.4822 - 4828, 2005/08
Graphite materials are used for structural components in High Temperature Gas-Cooled Reactor (HTGR) core because of their excellent thermo/mechanical properties. Thermal conductivity of graphite components is reduced by neutron irradiation in reactor operation. The reduced conductivity is expected to be recovered by thermal annealing effect when irradiated graphite component is heated above irradiated temperature. In the present study, temperature analyses considering the annealing effect of the HTGR core at a depressurization accident were carried out and influence of annealing effect on maximum fuel temperature was investigated. The analyses show that the annealing effect can reduce the fuel temperature about 100C at the maximum, and it is possible to evaluate the maximum fuel temperature more appropriately. It was also shown that the core-temperature of High Temperature Engineering Test Reactor (HTTR) at the safety demonstration tests can be analyzed with the developed evaluation method considering annealing effect.
Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki
Heat Transfer-Asian Research, 34(5), p.293 - 308, 2005/07
Water cooling panels have been adopted as the vessel cooling system of the HTTR to cool the reactor core indirectly by natural convection and thermal radiation. In order to investigate the heat transfer characteristics of high temperature gas in a vertical annular space between the reactor pressure vessel and cooling panels of the HTTR, we carried out experiments and numerical analyses on natural convection heat transfer coupled with thermal radiation heat transfer in an annulus between two vertical concentric cylinders with the inner cylinder heated and the outer cylinder cooled. In the present experiments, Rayleigh number based on the height of the annulus ranged from 2.0
10
to 5.410 for helium gas and from 1.210
to 3.510 for nitrogen gas. The numerical results were in good agreement with the experimental ones regarding the surface temperatures of the heating and cooling walls. As a result of the experiments and the numerical analyses, the heat transfer coefficient of natural convection coupled with thermal radiation was obtained.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
JAERI-Data/Code 2005-003, 31 Pages, 2005/06
JAERI-Data-Code-2005-003.pdf:4.83MB
Safety demonstration tests using the High Temperature engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. The TAC/BLOOST code was developed to analyze reactor and temperature transient during the coolant flow reduction test taking account of reactor dynamics. This paper describes the validation result of the TAC/BLOOST code with the measured values of gas circulators tripping tests at 30 % (9 MW). It was confirmed that the TAC/BLOOST code was able to analyze the reactor transient during the test.
Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Maeda, Yukimasa; Takeda, Tetsuaki; Nishihara, Tetsuo; Inagaki, Yoshiyuki
JAERI-Tech 2005-014, 89 Pages, 2005/03
JAERI-Tech-2005-014.pdf:7.25MB
In a hydrogen production system using HTTR, it is required to control a secondary helium gas temperature within an allowable value at an intermediate heat exchanger (IHX) inlet to prevent a reactor scram. To mitigate thermal disturbance of the secondary helium gas caused by the hydrogen production system, a cooling system of the secondary helium gas using a steam generator(SG) and a radiator will be installed at the downstream of the chemical reactor. In order to verify a numerical analysis code of the cooling system, numerical analysis has been conducted. The pressure controllability in SG is highly affected by the heat transfer characteristics of air which flows outside of the heat exchanger tube of the radiator. In order to verify a numerical analysis code of the cooling system, the heat transfer characteristics of air has been investigated with experimental results of a mock-up model test. It was confirmed that numerical analysis results were agreed well with experimental results, and the analysis code was successfully verified.
Ashikagaya, Yoshinobu; Kawasaki, Tomokatsu; Yoshino, Toshiaki; Ishida, Keiichi
JAERI-Tech 2005-010, 81 Pages, 2005/03
JAERI-Tech-2005-010.pdf:16.65MB
no abstracts in English
Inaba, Yoshitomo; Ohashi, Hirofumi; Nishihara, Tetsuo; Sato, Hiroyuki; Inagaki, Yoshiyuki; Takeda, Tetsuaki; Hayashi, Koji; Takada, Shoji
Nuclear Engineering and Design, 235(1), p.111 - 121, 2005/01
Times Cited Count：9 Percentile：52.13(Nuclear Science & Technology)Prior to the connection of a hydrogen production plant to the HTTR, the fluctuation tests of the chemical reaction in the steam reformer with the mock-up test facility of the HTTR hydrogen production system were carried out for the establishment and demonstration of the control technology. As a result, it was shown that the HTTR hydrogen production system with the same control system as the mock-up test facility can provide stable controllability for any disturbance at the steam reformer without the influence to the reactor. In addition, a dynamic simulation code for the HTTR hydrogen production system was verified with the obtained test data.
Ogawa, Masuro; Nishihara, Tetsuo
Nuclear Engineering and Design, 233(1-3), p.5 - 10, 2004/10
Times Cited Count：27 Percentile：82.91(Nuclear Science & Technology)An amount of primary energy supply in Japan is increasing year by year. Much energy such as oil, coal and natural gas is imported so that the self-sufficiency ratio in Japan is only 20 % even if including nuclear energy. An amount of energy consumption is also increasing especially in commercial and resident sector and transport sector. As a result, a large amount of greenhouse gas was emitted into the environment. Nuclear energy plays the important role in energy supply in Japan. Japan Atomic Energy Research Institute (JAERI) has been carried out research and development of a hydrogen production system using a high temperature gas cooled reactor (HTGR). The HTTR project aims at the establishment of the HTGR hydrogen production system. Reactor technology of the HTGR, hydrogen production technology with thermochemical water splitting process and system integration technology between the HTGR and a hydrogen production plant are developed in the HTTR project.
Takada, Eiji*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke
Nuclear Engineering and Design, 233(1-3), p.37 - 43, 2004/10
Times Cited Count：13 Percentile：62.98(Nuclear Science & Technology)The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88 % of the total flow is achieved at minimum. The maximum fuel temperature appears during the high temperature test operation, and reaches 1492 C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 C. It is confirmed that the core thermal-hydraulic design gives conservative results.
Sumita, Junya; Nakano, Masaaki*; Tsuji, Nobumasa*; Shibata, Taiju; Ishihara, Masahiro
JAERI-Tech 2004-055, 25 Pages, 2004/08
JAERI-Tech-2004-055.pdf:4.25MB
Neutron irradiation remarkably reduces the thermal conductivity of graphite, and the reduced thermal conductivity is recovered by annealing effect if the graphite is heated above the irradiation temperature. Therefore, it is expected that the reduced thermal conductivity of graphite components in the HTGR could be recovered by the annealing effect in accidents, such as a depressurization accident. Then, an analytical investigation of the annealing effect on thermal performance of a HTGR core was carried. The analysis showed that the annealing effect reduces the maximum fuel temperature about 70C, and it is important to introduce the annealing effect appropriately in the temperature analysis of the core components and reactor internals. In addition, an annealing test method was investigated to evaluate the effect quantitatively, and the test plan was made.
Inaba, Yoshitomo; Zhang, Y.*; Takeda, Tetsuaki; Shiina, Yasuaki
Nihon Kikai Gakkai Rombunshu, B, 70(694), p.1518 - 1525, 2004/06
no abstracts in English
Takamatsu, Kuniyoshi; Katanishi, Shoji; Nakagawa, Shigeaki; Kunitomi, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(1), p.76 - 87, 2004/03
The Gas Turbine High Temperature Reactor 300 (GTHTR300) composed of an inherent safe 600MWt reactor and a closed gas turbine power conversion system is a high efficient and economically competitive HTGR to be deployed in 2010s. To analyze the plant dynamics and the thermal hydraulics of the GTHTR300, a new analytical code (Conan-GTHTR) based on 'RELAP5/MOD3' has been developed and applied to heat transfer calculations of the High Temperature Engineering Test Reactor (HTTR) for its verification. The results proved that the new code was available for transient simulations in Higt Temperature Gas-Cooled Reactor systems.
