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Li, X.; Yamaji, Akifumi*; Sato, Ikken*; Yamashita, Takuya
Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Motome, Yuiko; Agake, Toshiki; Yanagisawa, Hiroshi
JAEA-Technology 2024-015, 30 Pages, 2025/01
JAEA-Technology-2024-015.pdf:1.36MB
The tables for calibration of control rods were verified, which is used positive period method and improved rod drop method of periodic inspection at Nuclear Safety Research Reactor (NSRR). Those tables are "DOUBLING TIME-REACTIVITY" and "DECAY OF NEUTRON FLUX AFTER INSTANTANEOUS REDUCTION OF REACTIVITY". They are prepared around 1975. Since those tables do not clearly express source of values and records of data used in calculations, the authors verified those tables again. For the verification, the tables were reproduced as follows. For the positive period method, the relationship between the period and reactivity was analytically evaluated by using the inhour equation with NSRR's parameters. For the improved rod drop method, the ratios of neutron flux after the rod drop with parameters of negative reactivities was calculated using the EUREKA- 2 code. As a result, the values described in the tables well agree with those by the present evaluation because it is confirmed that standard deviations of the differences in the value by between the present evaluation and the tables are less than 0.035%. For this reason, it is verified that these tables are valid in the practical use for NSRR operations.
Fujita, Tatsuya
Proceedings of International Conference on Physics of Reactors (PHYSOR 2024) (Internet), p.718 - 727, 2024/04
The convergence process of the k-infinity uncertainty during random-sampling-based uncertainty quantification was compared between several efficient sampling techniques. The k-infinity uncertainty was evaluated by statistically processing several times of SERPENT 2.2.1 calculations using perturbed ACE files based on JENDL-5 cross-section covariance data. The antithetic sampling (AS), the Latin hypercube sampling (LHS), the control variates (CV), and the combination approaches of them were focused on in the present paper. In PWR-UO
fuel assembly geometry without the nuclide depletion, as discussed in past studies, AS and LHS showed higher efficient convergence than nominal sampling without any efficient sampling techniques. In terms of CV, though a stand-alone application did not have a large impact on the k-infinity uncertainty convergence, its performance was improved in combination with AS, as discussed in the past study. In addition, a new combined approach of LHS and CV (CV+LHS) was proposed in the present paper. CV+LHS improved the k-infinity uncertainty convergence and was more efficient than CV+AS. The main reason for this improvement was that the convergence for the mean value of alternative parameters in CV was enhanced by applying LHS. Consequently, this study proposed the new combined approach of CV+LHS and confirmed its efficiency performance for the random-sampling-based uncertainty quantification in the PWR-UO fuel assembly geometry. The applicability of CV+LHS for the nuclide-depletion calculations will be confirmed in future studies.
Morimoto, Kyoichi; Ono, Takahiro; Kakutani, Satomi; Yoshida, Moeka; Suzuki, Soichiro
Journal of Robotics and Mechatronics, 36(1), p.125 - 133, 2024/02
The Naraha Center for Remote Control Technology Development was established for the purpose of developing and verifying remote control equipment for promoting the decommissioning of the Fukushima Daiichi Nuclear Power Station and the external use of this center was started in 2016. The mission of this center is to contribute to the decommissioning of the Fukushima Daiichi Nuclear Power Station and for the reconstruction of Fukushima Prefecture. In this review, we describe the equipment related to the full-scale mock-up test, the component test for a remote-control device and the virtual reality system in this center. In addition, the case examples for usage of these equipment are introduced.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10
The benchmark analyses on the TRIGA reactor of the IEU-COMP-THERM-013 (ICT-013) in the ICSBEP handbook using the MVP version 3 code were carried out with the Japanese, US and European nuclear data libraries, including JENDL-5. The analyses on effective neutron multiplication factors (k
's) were also performed for another TRIGA reactor defined in the ICT-003 as well as the ICT-013. As a result, it is confirmed that the calculated k's vary in the range of 0.8% with the libraries. The results also suggest that there might be unknown bias in the ICT-013 for the calculated k's. For the analyses on the control rod worths of the ICT-013, it is confirmed that differences in the worths among the libraries become smaller than those in k's. Most of the errors involved in k's are considered to be cancelled in the calculation of the worths since the worths are obtained by subtraction of the reciprocals of two sorts of k's. The differences in the worths by between the benchmark and alternative methods defined in the ICT-013 were tried to be examined from the aspect of the differences in horizontal flux distributions by between those methods. It is confirmed that the differences in the worths for two shim control rods, which are fully withdrawn at a delayed critical state, are well understood by that attempt, but on the other hand the differences are not sufficiently explained for a regulating control rod, which is partially inserted to attain delayed criticality.
Yamada, Taichi; Watanabe, Kaho; Suzuki, Soichiro; Kawabata, Kuniaki
Automation Systems, 39(464), p.88 - 92, 2023/09
In Fukushima Daiichi Nuclear Power Station (FDNPS) emergency response and decommissioning, high radiation or unknown environments significantly restrict human workers' activity. Thus, a remotely controlled robot is essential to operate in such an environment instead of human workers. However, remote robot operation is not easy, and it is required to understand the robot's capability, that is, what/how the robot can do on the site. Therefore, robot evaluation method development is important for remote robot operation in disaster sites. We survey the required capabilities for a remotely controlled robot from the remote operation cases in FDNPS and develop test methods to evaluate the capabilities. This paper introduces the survey of FDNPS remote operation cases and the test method development.
Yamada, Taichi; Watanabe, Kaho; Suzuki, Soichiro; Kawabata, Kuniaki
Keisoku To Seigyo, 62(5), p.268 - 271, 2023/05
In Fukushima Daiichi Nuclear Power Station (FDNPS) emergency response and decommissioning, high radiation or unknown environments significantly restrict human workers' activity. Thus, a remotely controlled robot is essential to operate in such an environment instead of human workers. However, remote robot operation is not easy, and it is required to understand the robot's capability, that is, what/how the robot can do on the site. Therefore, robot evaluation method development is important for remote robot operation in disaster sites. We survey the required capabilities for a remotely controlled robot from the remote operation cases in FDNPS and develop test methods to evaluate the capabilities. This paper introduces the survey of FDNPS remote operation cases and the test method development.
Takabe, Yugo; Otsuka, Noriaki; Fuyushima, Takumi; Sayato, Natsuki; Inoue, Shuichi; Morita, Hisashi; Jaroszewicz, J.*; Migdal, M.*; Onuma, Yuichi; Tobita, Masahiro*; et al.
JAEA-Technology 2022-040, 45 Pages, 2023/03
JAEA-Technology-2022-040.pdf:6.61MB
Because of the decommission of the Japan Materials Testing Reactor (JMTR), the domestic neutron irradiation facility, which had played a central role in the development of innovative nuclear reactors and the development of technologies to further improve the safety, reliability, and efficiency of light water reactors, was lost. Therefore, it has become difficult to pass on the operation techniques of the irradiation test reactors and irradiation technologies, and to train human resources. In order to cope with these issues, we conducted a study on the implementation of irradiation tests using overseas reactors as neutron irradiation sites as an alternative method. Based on the "Arrangement between the National Centre for Nuclear Research and the Japan Atomic Energy Agency for Cooperation in Research and Development on Testing Reactor," the feasibility of conducting an irradiation test at the MARIA reactor (30 MW) owned by the National Centre for Nuclear Research (NCBJ) using the temperature control system, which is one of the JMTR irradiation technologies, was examined. As a result, it was found that the irradiation test was possible by modifying the ready-made capsule manufactured in accordance with the design and manufacturing standards of the JMTR. After the modification, a penetration test, an insulation continuity test, and an operation test in the range of room temperature to 300
C, which is the operating temperature of the capsule, were conducted and favorable results were obtained. We have completed the preparations prior to transport to the MARIA reactor.
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2022-030, 80 Pages, 2023/02
JAEA-Technology-2022-030.pdf:2.57MBJAEA-Technology-2022-030(errata).pdf:0.11MB
Nuclear criticality benchmark analyses were carried out for TRIGA-type reactor systems in which uranium-zirconium hydride fuel rods are loaded by using the continuous-energy Monte Carlo code MVP with the evaluated nuclear data library JENDL-5. The analyses cover two sorts of benchmark data, the IEU-COMP-THERM-003 and IEU-COMP-THERM-013 in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and effective neutron multiplication factors, reactivity worths for control rods etc. were calculated by JENDL-5 in comparison with those by the previous version of JENDL. As the results, it was confirmed that the effective neutron multiplication factors obtained by JENDL-5 were 0.4 to 0.6% greater than those by JENDL-4.0, and that there were no significant differences in the calculated reactivity worths by between JENDL-5 and JENDL-4.0. Those results are considered to be helpful for the confirmation of calculation accuracy in the analyses on NSRR control rod worths, which are planned in the future.
Kawamura, Sho; Kikuchi, Masanobu; Hosoya, Toshiaki
JAEA-Technology 2021-041, 103 Pages, 2023/02
JAEA-Technology-2021-041.pdf:8.7MB
In response to new regulatory standard for research and test reactor which is enforced December 2013, JRR-3 got license in November 2018 by formulate new design basis ground motion. After that we evaluated for insertion property of control rod using that new design basis ground motion, and that evaluation results were accepted as approval of the design and construction method by Nuclear Regulation Authority. Now, we re-evaluated to insertion property of control rod about neutron absorber and follower fuel element by time history response analysis method. In this report, it shows that new results have sufficiency of margin compared with the past results that are accepted as approval of the design and construction method.
-ray discrimination based on the property and thickness controls of scintillators using Li glass and LiCAF(Ce) in a -ray fieldKaburagi, Masaaki; Shimazoe, Kenji*; Terasaka, Yuta; Tomita, Hideki*; Yoshihashi, Sachiko*; Yamazaki, Atsushi*; Uritani, Akira*; Takahashi, Hiroyuki*
Nuclear Instruments and Methods in Physics Research A, 1046, p.167636_1 - 167636_8, 2023/01
Times Cited Count：7 Percentile：91.75(Instruments & Instrumentation)We focus on the thickness and property controls of inorganic scintillators used for thermal neutron detection in intense -ray fields without considering pulse shape discrimination techniques. GS20
(a lithium glass) and LiCaAlF
:Ce(LiCAF:Ce) cintillators with thicknesses of 0.5 and 1.0 mm, respectively, have been employed. Pulse signals generated by photomultiplier tubes, to which the scintillators were coupled, were inserted into a digital pulse processing unit with 1 Gsps, and the areas of waveforms were integrated for 360 ns. In a 
Co -ray field, the neutron detection for GS20 with a 0.5-mm thickness was possible at dose rates of up to 0.919 Gy/h; however, for LiCAF:Ce, neutron detection was possible at 0.473 Gy/h, and it failed at 0.709 Gy/h. Threfore, in a Co -ray field, the neutron/-ray discrimination of GS20 was better than that of LiCAF:Ce due to its better energy resolution and higher detection efficiency.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Akaev, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-13) (Internet), 12 Pages, 2022/09
The In-Vessel Retention (IVR) of molten-core in Core Disruptive Accidents (CDAs) is of prime importance in enhancing the safety of sodium-cooled fast reactors. One of the main subjects in ensuring IVR is to design the Control Rod Guide Tube (CRGT) which allows effective discharge of molten core materials from the core region. The effectiveness of the CRGT design is assessed through CDA analyses, and it is reasonable for these analyses to develop a computer code collaborated with experimental researches. Thus, experiments addressing the discharge behavior of the molten-core materials through the CRGT have proceeded as one of the subjects in the collaboration research named the EAGLE-3 project, and the obtained experimental results are reflected in the development of the SIMMER code. In this project, a series of out-of-pile tests using molten-alumina as the fuel simulant was conducted to understand the discharge behavior of molten-core materials through the CRGT. In this study, in order to investigate the effect of an internal structure in the CRGT on the discharge behavior of the molten-core materials, the data of an out-of-pile test in which the molten-alumina penetrated to a duct with the internal structure were analyzed. In addition, the post-test analysis using the SIMMER code was conducted and the results were compared with the test results.
Isogawa, Hiroki*; Naoi, Motomasa*; Yamasaki, Seiji*; Ho, H. Q.; Katayama, Kazunari*; Matsuura, Hideaki*; Fujimoto, Nozomu*; Ishitsuka, Etsuo
JAEA-Technology 2022-015, 18 Pages, 2022/07
JAEA-Technology-2022-015.pdf:1.37MB
As a summer holiday practical training 2021, the impact of 10 years long-term shutdown on critical control rod position of the HTTR and the delayed neutron fraction (
) of the VHTRC-1 core were investigated using Monte-Carlo MVP code. As a result, a long-term shutdown of 10 years caused the critical control rods of the HTTR to withdraw about 4.0
0.8 cm compared to 3.9 cm in the experiment. The change in critical control rods position of the HTTR is due to the change of some fission products such as 
Pu, Am, 
Pm, Sm, 
Gd. Regarding the calculation of the VHTRC-1 core, the value is underestimate of about 10% in comparison with the experiment value.
Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
Transactions of the 26th International Conference on Structural Mechanics in Reactor Technology (SMiRT-26) (Internet), 8 Pages, 2022/07
Utilizing fracture control, we are developing a technology to suppress the expansion of damage caused by an event that exceeds the design assumption. We made a plan to develop three issues; (1) Technology for mitigating failure consequence at extremely high temperatures, (2) Technology for mitigating failure consequence against excessive earthquakes, and (3) Methodology for improving reactor structure resilience.
Kamiyama, Kenji; Matsuba, Kenichi; Kato, Shinya; Imaizumi, Yuya; Mukhamedov, N.*; Akayev, A.*; Pakhnits, A.*; Vurim, A.*; Baklanov, V.*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi*
Nuclear Technology, 208(3), p.484 - 493, 2022/03
Times Cited Count：2 Percentile：22.08(Nuclear Science & Technology)An Ag-In-Cd control rod alloy was heated in argon or oxygen at 1073-1673 K for 60-3600 s and the release behavior of the elements was examined. Complete liquefaction of the alloy occurred between 1123 and 1173 K, and elemental release was quite limited below the liquefaction temperature. In argon, almost all of the Cd content was released within 3600 s at 
1173 K and within 60 s at 1573 K, while the released fractions of Ag and In were 
3% and 8%, respectively. In oxygen, the release of Cd, which was quite small at temperatures up to 1573 K, drastically increased to 
30-50% at 1673 K for short periods. Releases of Ag and In were also small in oxygen under the examined conditions. Comparison with the experimental data suggests that conventional empirical release models may underestimate the Cd release at lower temperatures just after control rod failure in severe accidents.
Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo
Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11
Times Cited Count：1 Percentile：10.83(Nuclear Science & Technology)Ho, H. Q.; Fujimoto, Nozomu*; Hamamoto, Shimpei; Nagasumi, Satoru; Goto, Minoru; Ishitsuka, Etsuo
Nuclear Engineering and Design, 377, p.111161_1 - 111161_9, 2021/06
Times Cited Count：3 Percentile：33.11(Nuclear Science & Technology)Tsubaki, Hirohiko; Koizumi, Satoshi*
JAEA-Technology 2020-016, 16 Pages, 2020/11
JAEA-Technology-2020-016.pdf:2.96MB
Maintenance and Operation Section for Remote Control Equipment in Naraha Center for Remote Control Technology Development is the main part of the nuclear emergency response team of JAEA deal with Act on Special Measures Concerning Nuclear Emergency Preparedness. The section needs to train operators from every nuclear facility in JAEA to control crawler-type robots, and so on. A driving training of a crawler-type robot used a reciprocating passage (U-shaped passage look from above) is one of the important training programs. The section always assembled a reciprocating passage with borrowed parts from other sections for every training of being used the passage. The section designed and produced training-way system included a reciprocating passage with stairs in 2019 fiscal year. The system makes the section members labor-saving, possible to set any time for training and diverse training-ways with easy assembling system. This report shows design and produce training-way system for crawler-type robots against nuclear emergency of JAEA facilities by Maintenance and Operation Section for Remote Control Equipment.
Kokusen, Junya; Akasaka, Shingo*; Shimizu, Osamu; Kanazawa, Hiroyuki; Honda, Junichi; Harada, Katsuya; Okamoto, Hisato
JAEA-Technology 2020-011, 70 Pages, 2020/10
JAEA-Technology-2020-011.pdf:3.37MB
The Uranium Enrichment Laboratory in the Japan Atomic Energy Agency (JAEA) was constructed in 1972 for the purpose of uranium enrichment research. The smoke emitting accident on 1989 and the fire accident on 1997 had been happened in this facility. The research on uranium enrichment was completed in JFY1998. The decommissioning work was started including the transfer of the nuclear fuel material to the other facility in JFY2012. The decommissioning work was completed in JFY2019 which are consisting of removing the hood, dismantlement of wall and ceiling with contamination caused by fire accident. The releasing the controlled area was performed after the confirmation of any contamination is not remained in the target area. The radioactive waste was generated while decommissioning, burnable and non-flammable are 1.7t and 69.5t respectively. The Laboratory will be used as a general facility for cold experiments.
