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Nagasumi, Satoru; Hasegawa, Toshinari; Iigaki, Kazuhiko; Nakagawa, Shigeaki; Kubo, Shinji; Shimazaki, Yosuke; Nakajima, Kunihiro; Sakurai, Yosuke; Shinohara, Masanori; Saito, Kenji; et al.
Nuclear Engineering and Design, 446, p.114542_1 - 114542_14, 2026/01
To demonstrate HTGR's safety features, a loss-of-forced-cooling (LOFC) test was conducted using the HTTR. In this test, the forced cooling in the reactor core was intentionally lost by shutting down all helium gas circulators (HGCs) without reactor scram. During steady-state operation at 100% reactor power (30 MW), after the LOFC, the reactor power spontaneously decreased. This power reduction occurred due to the negative reactivity feedback effect triggered by an increase in core temperature. The power stabilized at a lower value of 1.2% after re-criticality. Additionally, the measured radioactivity concentration in the primary coolant remained nearly unchanged during this LOFC operation and during an immediately subsequent HTTR operation. This indicates no failure of the coated particle fuel, even after the increase in core temperature associated with the LOFC event. These results provide experimental evidence of the safety features of HTGRs.
Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09
Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.47 - 51, 2025/09
Bachmann, A. M.*; Nishihara, Kenji; Richards, S.*; Abe, Takumi; Feng, B.*
Progress in Nuclear Science and Technology (Internet), 8, p.11 - 16, 2025/09
Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.
JAEA-Research 2025-005, 23 Pages, 2025/07
JAEA-Research-2025-005.pdf:2.68MB
A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.
Abe, Takumi; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*
Progress in Nuclear Science and Technology (Internet), 7, p.299 - 304, 2025/03
Currently, much research continues on stable energy sources that do not emit CO
in order to achieve a carbon-neutral and sustainable society. Nuclear energy is one of the such sources, and various new reactors and reprocessing technologies are being developed. In order to implement the nuclear fuel cycle with these technologies, a nuclear fuel cycle simulator is required to quantitatively evaluate various quantities, such as the distribution of nuclear fuel materials and the scale of waste loading. For this purpose, NMB4.0 was developed in collaboration with Tokyo Institute of Technology and Japan Atomic Energy Agency. This code calculates the material balance of 179 nuclides including actinides and fission products (FPs) from the front-end to the back-end and simulates the nuclear fuel cycle in an integrated manner. Unlike other nuclear fuel cycle simulators, the code is capable of performing precise back-end analyses such as the number of radioactive wastes and the scale of the geological repository considering heat generation of waste package under diverse nuclear energy scenario, and is an open source code that runs on Microsoft Excel. By these features, it is possible to quantitatively study nuclear energy utilization strategies with various stakeholders. The presentation will detail the numerical model used in NMB4.0.
Nishihara, Kenji
Enerugi, Shigen, 45(6), p.359 - 363, 2024/11
no abstracts in English
Nishihara, Kenji; Sugawara, Takanori; Fukushima, Masahiro; Iwamoto, Hiroki; Katano, Ryota; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
A pilot plant for the accelerator-driven system is proposed as a scaled-down version of a lead-bismuth cooled ADS with 800 MW thermal output for transmutation of minor actinides. In this presentation, the design policy of the pilot plant is presented.
Bachmann, A. M.*; Richards, S.*; Feng, B.*; Nishihara, Kenji; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
This work demonstrates the value of code verification as an initial step in utilizing fuel cycle simulation. Cyclus and NMB are open-source fuel cycle simulators that provide computational modeling of nuclear fuel cycle alternatives and were chosen by Argonne National Laboratory and Japan Atomic Energy Agency (JAEA), respectively, for a multi-year collaboration on fuel cycle benchmarks. Both are relatively new and can be improved after conducting a rigorous code-to-code comparison. Initial verification of these simulators was performed using a set of hypothetical scenarios for once-through and multi-recycle fuel cycles. The results of this work identify how differences in scenario definitions and the modeling methodologies of the two simulators lead to differences in results in material inventories, mass flows, and other important metrics for fuel cycle assessments.
Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Nishihara, Kenji; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 2 Pages, 2024/10
Using the dynamic nuclear fuel cycle simulator NMB4.0, the mass balance analysis of the nuclear fuel cycle assuming the introduction of the metal fuel fast reactor in the second half of this century was evaluated. The impact of the introduction of the fast reactor cycle on the back-end including final disposal was discussed.
Abe, Takumi; Nishihara, Kenji
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
The robustness of whole of a nuclear fuel cycle (NFC) can be evaluated by a simulation of future operation factors (OFs) of the NFC facilities and mass flow analysis coupled with the simulated OF. In this study, the impact of a reprocessing plant OF on a fast reactor OF was quantified.
Abe, Takumi; Nishihara, Kenji
Journal of Nuclear Science and Technology, 61(8), p.1048 - 1060, 2024/08
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Methods of direct disposal of mixed oxide spent fuel (MOXSF) consumed in light water reactors were investigated via heat transfer calculation. The temperature of the buffer material surrounding the waste is the most stringent limitation on the direct disposal of MOXSF. Therefore, the effects on the maximum temperature of the buffer material were examined by changing the occupied area, cooling term of MOXSF, and other parameters considering the direct disposal of uranium spent fuel. The results showed that it is necessary to change the cooling term and disposal depth in addition to the change of the occupied area. Accordingly, some disposal concepts that satisfy the limitation of the maximum buffer material temperature were derived, and estimates revealed that the occupied area per unit waste of the MOXSF is three to five times that of uranium spent fuel.
Yamagishi, Isao; Hato, Shinji*; Nishihara, Kenji; Tsubata, Yasuhiro; Sagawa, Yusuke*
JAEA-Data/Code 2024-002, 63 Pages, 2024/07
JAEA-Data-Code-2024-002.pdf:2.91MB
JAEA-Data-Code-2024-002-appendix(CD-ROM).zip:9.42MB
Adsorption columns filled with zeolite are used to treat contaminated water containing radioactive cesium generated by the Fukushima Daiichi Nuclear Power Station accident. As the contaminated water treatment progresses, the radioactive cesium in the adsorption column becomes highly concentrated, and the adsorption column becomes a high radiation source. To evaluate the radiation effects such as decay heat and radiolytic hydrogen production in the adsorption column, the concentration of radioactive cesium in the adsorption column is necessary, but since it is difficult to evaluate the concentration by measurement, it is estimated by simulation. In this research, a zeolite column adsorption dynamics simulation (Zeolite Adsorption Column: ZAC) code was developed to calculate the concentration of radioactive materials such as radioactive cesium in a zeolite filled adsorption column when they are injected into the column. The code was validated through comparison of calculation results with existing codes and experimental results of small column tests. This report presents the details of the model, the handling of the code, and the validity of the results for the developed code.
Nishihara, Kenji
Genshiryoku Nenkan 2024, p.78 - 83, 2023/10
The aim and overall picture of partitioning and transmutation technology will be described, and the recent research and development status of JAEA for each of partitioning and transmutation will be outlined.
Watanabe, Nao; Yamashita, Susumu; Uesawa, Shinichiro; Nishihara, Kenji; Yoshida, Hiroyuki
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.3522 - 3534, 2023/08
Accelerator-driven system (ADS), the coolant of which is lead-bismuth eutectic (LBE), has been designed by Japan Atomic Energy Agency. Estimating corrosion rate at the wall surface of LBE channel is an important issue in considering safety and the life of the entire structure. The corrosion rate depends on state of oxygen layers forming at the material surface. Therefore, this study aims to develop a method to evaluate the corrosion rate in ADS for the design study by estimation of the oxide layer growth and dissolution (OLGD) rates by means of numerical analysis. The OLGD rates, mass transfer rates of oxygen and iron between the material and LBE and advection-diffusion rates of them in LBE depend on each other. Therefore, in order to estimate OLGD rates, the three numerical analysis models should be coupled. For the advection-diffusion calculation, to use CFD code should be reasonable approach to analyze complex flow in ADS, while for the OLGD and the mass transfer calculation, to use some correlation equations should be reasonable because their scales are much smaller than the advection-diffusion. The present work has developed the analysis method of OLGD rates by using JUPITER code, which is CFD code developed in JAEA. In terms of the correlation equations of OLGD and mass transfer rates, existing models used in a previous study were used with modified.
Okamura, Tomohiro*; Katano, Ryota; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*
Journal of Nuclear Science and Technology, 60(6), p.632 - 641, 2023/06
Times Cited Count：3 Percentile：35.97(Nuclear Science & Technology)The Okamura explicit method (OEM) for depletion calculation was developed by modifying the matrix exponential method for dynamic nuclear fuel cycle simulation. The OEM suppressed the divergence of the calculation for short half-life nuclides, even for long time steps. The computational cost of the OEM was small, equivalent to the Euler method, and it maintained sufficient accuracy for the fuel cycle simulation.
Nishihara, Kenji
JAEA-Conf 2022-001, p.63 - 67, 2022/11
This tutorial contains a role of accelerator-driven system (ADS) in the nuclear fuel cycle and necessity of nuclear data to realize the ADS. After an overview of Japanese nuclear fuel cycle and government direction, geological disposal concept of high-level waste (HLW) will be described. By partitioning problematic elements from the HLW and transmuting, utilizing or storing them, geological disposal can be changed. ADS plays a role of transmuting minor actinide (MA) separated from HLW to fission product (FP), which are less radio-toxic than MA. The principle of ADS will be introduced with technological issues, and finally utilization of nuclear data for R&D on ADS will be introduced.
Sato, Rika*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Hayashi, Hirokazu; Sugawara, Takanori; Nishihara, Kenji
Dai-43-Kai Nihon Netsu Bussei Shimpojiumu Koen Rombunshu (CD-ROM), 3 Pages, 2022/10
no abstracts in English
Nishihara, Kenji
Genshiryoku Nenkan 2023, p.78 - 83, 2022/10
The overall picture and the goals of partitioning and transmutation technology are described. Then, the main technologies currently being developed for partitioning and transmutation are respectively outlined. Finally, a comprehensive evaluation of the nuclear fuel cycle including partitioning and transmutation, conducted in the Atomic Energy Society of Japan is introduced.
