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Fujimoto, Nozomu*; Fukuda, Kodai*; Honda, Yuki*; Tochio, Daisuke; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo
JAEA-Technology 2021-008, 23 Pages, 2021/06
JAEA-Technology-2021-008.pdf:2.62MB
The effect of mesh division around the burnable poison rod on the burnup calculation of the HTTR core was investigated using the SRAC code system. As a result, the mesh division inside the burnable poison rod does not have a large effect on the burnup calculation, and the effective multiplication factor is closer to the measured value than the conventional calculation by dividing the graphite region around the burnable poison rod into a mesh. It became clear that the mesh division of the graphite region around the burnable poison rod is important for more appropriately evaluating the burnup behavior of the HTTR core..
Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021902_1 - 021902_6, 2020/04
Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Fujimoto, Nozomu*; Ishitsuka, Etsuo
Applied Radiation and Isotopes, 140, p.209 - 214, 2018/10
Times Cited Count：3 Percentile：30.05(Chemistry, Inorganic & Nuclear)Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
Takada, Shoji; Honda, Yuki*; Inaba, Yoshitomo; Sekita, Kenji; Nemoto, Takahiro; Tochio, Daisuke; Ishii, Toshiaki; Sato, Hiroyuki; Nakagawa, Shigeaki; Sawa, Kazuhiro*
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 7 Pages, 2018/10
Nuclear heat utilization systems connected to HTGRs will be designed on the basis of non-nuclear grade standards for easy entry of chemical plant companies, requiring reactor operations to continue even if abnormal events occur in the systems. The inventory control is considered as one of candidate methods to control reactor power for load following operation for siting close to demand area, in which the primary gas pressure is varied while keeping the reactor inlet and outlet coolant temperatures constant. Numerical investigation was carried out based on the results of nuclear heat supply fluctuation tests using HTTR by non-nuclear heating operation to focus on the temperature transient of the reactor core bottom structure by imposing stepwise fluctuation on the reactor inlet temperature under different primary gas pressures below 120C. As a result, it was emerged that the fluctuation absorption characteristics are not deteriorated by lowering pressure. It was also emerged that the reactor outlet temperature did not reach the scram level by increasing the reactor inlet temperature 10 C stepwise at 80% of the rated power as same with the full power case.
Honda, Yuki; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi
Journal of Nuclear Engineering and Radiation Science, 4(3), p.031013_1 - 031013_11, 2018/07
There is growing interest in uncertainty analysis for probabilistic risk assessment (PRA). The focus of this research is to propose and trial investigate the new approach which identify influencing factors for uncertainty in a systematic manner for High Temperature Gas -cooled Reactor (HTGR). As a trial investigation, this approach is tested to evaluation of maximum fuel temperature in a depressurized loss-of-forced circulation (DLOFC) accident and failure of mitigation systems such as control rod systems from the view point of reactor dynamics and thermal hydraulic characteristics. As a result, 16 influencing factors are successfully selected in accordance with the suggested procedure. In the future, the selected influencing factors will be used as input parameter for uncertainty propagation analysis.
Ho, H. Q.; Honda, Yuki; Motoyama, Mizuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Ishitsuka, Etsuo
Applied Radiation and Isotopes, 135, p.12 - 18, 2018/05
Times Cited Count：7 Percentile：58.07(Chemistry, Inorganic & Nuclear)Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji
Annals of Nuclear Energy, 112, p.42 - 47, 2018/02
Times Cited Count：8 Percentile：62.99(Nuclear Science & Technology)Honda, Yuki; Sato, Hiroyuki; Nakagawa, Shigeaki; Ohashi, Hirofumi
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 9 Pages, 2017/07
There is growing interest in uncertainty analysis for probabilistic risk assessment (PRA). Our target is the uncertainty analysis method development for depressurized loss-of-forced circulation (DLOFC) accident with failure of control rod systems (CRS). As one of key elements, this paper focuses on the quantification of uncertainty for the fuel temperature which is dominant for a source term analysis. As an initial step, this paper aims to suggest a procedure to identify influencing factors which is input parameter for uncertainty analysis, and shows the results of derivation of variable parameters by expansion of dynamic equation and extraction of uncertainties in variable factors.
Honda, Yuki; Inaba, Yoshitomo; Nakagawa, Shigeaki; Yamazaki, Kazunori; Kobayashi, Shoichi; Aono, Tetsuya; Shibata, Taiju; Ishitsuka, Etsuo
JAEA-Technology 2017-013, 20 Pages, 2017/06
JAEA-Technology-2017-013.pdf:2.52MB
Decay heat is one of an important factor for a safety evaluation of depressurized loss-of-forced cooling accident, a representative high consequence accident, in high temperature gas-cooled reactor (HTGR). Traditionally, a conservative decay heat curve is used for safety analysis according to the regulatory standards. On the other hand, there is growing interest in obtaining test data related to decay heat for the use of uncertainty analysis. However, such data has not been obtained for prismatic-type HTGR. Therefore, we have launched a test program to obtain the decay heat data from the HTTR. As an initial step, an applicability confirmation test of decay heat evaluation method for HTGR was conducted in February 2017 without non-nuclear heating condition. This report introduces an estimation method for the decay heat based on test data using HTTR and shows the results of validation of the reactor residual heat evaluation method which will be used to obtain the decay heat data based on test data.
Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji
Annals of Nuclear Energy, 103, p.114 - 121, 2017/05
Times Cited Count：8 Percentile：61.27(Nuclear Science & Technology)Ho, H. Q.; Morita, Keisuke*; Honda, Yuki; Fujimoto, Nozomu*; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Ono, Masato; Fujiwara, Yusuke; Honda, Yuki; Sato, Hiroyuki; Shimazaki, Yosuke; Tochio, Daisuke; Homma, Fumitaka; Sawahata, Hiroaki; Iigaki, Kazuhiko; Takada, Shoji
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04
Japan Atomic Energy Agency (JAEA) has carried out research and developments towards nuclear heat utilization of High Temperature Gas-cooled Reactor (HTGR) using High Temperature Engineering Test Reactor (HTTR). The nuclear heat utilization systems connected to HTGR will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations continue even if abnormal events occur in the systems. Heat application system abnormal simulating test with HTTR was carried out in non-nuclear heating operation to focus on the thermal effect in order to obtain data of the transient temperature behavior of the metallic components in the Intermediate Heat Exchanger (IHX). The IHX is the key components to connect the HTTR with the heat application system. In the test, the coolant helium gas temperature was heated up to 120
C by the compression heat of the gas circulators in the HTTR under the ideal condition to focus on the heat transfer. The tests were conducted by decreasing the helium gas temperature stepwise by increasing the mass flow rate to the air cooler. The temperature responses of the IHX were investigated. For the components such as the heat transfer tubes and heat transfer enhancement plates of IHX, the temperature response was slower in the lower position in comparison with the higher position. The reason is considered that thermal load fluctuation is imposed in the secondary helium gas which flows from the top to the bottom in the heat transfer tubes of the IHX. The test data are useful to verify the numerical model of the safety evaluation code.
Honda, Yuki; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
We have been conducting a source term evaluation method development for high temperature gas-cooled reactors considering structural failures in the major components. This paper present the results of transient analysis for depressurized loss-of-forced cooling accident with ruptures of the cross cut ducts and standpipe, which may be initiated by earthquake. The sequences accounts failures of mitigation systems such as core heat removal by Vessel Cooling System (VCS) and reactor shut down by control rod systems. We will show the effect of mitigation system failure to depressurized loss-of-forced cooling accident in the view point of fuel temperature and natural circulation flow rate which is important for source term evaluation. The major findings obtained in this study showed that multiple failures in mitigation systems for a representative HTGR plant do not aggravate the accident. The result demonstrated that a simplification of event sequence analysis and source term analysis can be achieved with a design fully utilizing the safety characteristics of HTGR.
Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji
JAEA-Technology 2016-040, 16 Pages, 2017/03
JAEA-Technology-2016-040.pdf:2.89MB
To study the packing effects of the truncated coated fuel particle on the criticality for the High Temperature engineering Test Reactor (HTTR), four alternative models including the truncated uniform model, the non-truncated uniform model, the truncated random model and the non-truncated random model for the arrangement of CFP in fuel compact were used, and the neutronic and criticality calculation were performed by using Monte Carlo MCNP6 code with ENDF/B-VII.1 cross section library. The results showed that the infinite multiplication factors (k
) in the truncated models were lower than those of the non-truncated models regardless of the uniform or random arrangement, and the four factors in the four-factor-formula showed that the difference of k was mainly attributed to the resonance escape probability. The difference in resonance escape probability is caused by the increase of capture reactions in the resonance region as the influence of spatial-self-shielding-effect. It is because the equivalent kernel diameter of the CFP for the truncated model is smaller than that of the non-truncated model.
Honda, Yuki; Fujimoto, Nozomu*; Sawahata, Hiroaki; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 3(1), p.011013_1 - 011013_4, 2017/01
The operating data of the HTTR with burn-up is very important for developments of HTGRs. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate of calculation code. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle. The temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.
Honda, Yuki; Fujimoto, Nozomu*; Sawahata, Hiroaki; Takada, Shoji; Sawa, Kazuhiro
Journal of Nuclear Engineering and Radiation Science, 3(1), p.011005_1 - 011005_6, 2017/01
In the HTTR, a two-step control rods insertion method for reactor scram is adopted. In the method, control rods at reflector region are inserted at the scram signal is initiated. The core should keep its subcriticality by reflector region control rods. Therefore, precise evaluation of control rods reactivity worth for reflector region is necessary. However, all cross section of control rods has been prepared for control rod in fuel region because the reactivity value of control rods in the fuel region is larger than that of control rods in the reflector region. This paper proposed the revised method of preparing the control rod cross section for first step control rod in reflector region.
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：1 Percentile：10.62(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.
Honda, Yuki; Fukaya, Yuji; Nakagawa, Shigeaki; Baker, R. I.*; Sato, Hiroyuki
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.704 - 713, 2016/11
A high-temperature gas-cooled reactor (HTGR) has superior safety characteristics. A loss of forced cooling (LOFC) test using a high-temperature engineering test reactor (HTTR) has been carried out to verify the inherent safety of an HTGR when forced cooling is diminished without reactor scram. In the test, an all-gas circulator was tripped with an initial reactor power of 9 MW and re-criticality was shown. This study focuses on developing a point kinetics method with RELAP5-3D code for an LOFC accident. There is a large temperature difference between the inlet and outlet of the core in an HTGR, and the temperature fluctuation range has been large in several accidents. We analyze the temperature dependency of xenon-135 reactivity and show that the temperature dependency of xenon-135 microscopic absorption cross-section affected the re-criticality time of the LOFC test.
Inaba, Yoshitomo; Honda, Yuki; Nishihara, Tetsuo
Proceedings of 8th International Topical Meeting on High Temperature Reactor Technology (HTR 2016) (CD-ROM), p.985 - 990, 2016/11
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 the normal operation is to be lower than the 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 and neutron fluence distributions, and core coolant flow distribution. The core coolant flow distribution calculation code used in the design stage of High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and not user-friendly. Therefore, a new core coolant flow distribution calculation code with a user-friendly system such as simple and easy operations and execution procedures has been developed. In this paper, the outline of the new code is described and the simulation result of an out-of-pile test with one fuel column is shown as the first step of the code validation. The simulation results provide good agreement with the test one.
