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Sasaki, Yuto*; Sano, Aaru; Sasaki, Shinji; Iwamoto, Nobuyuki; Ouchi, Kazuki; Kitatsuji, Yoshihiro; Takaki, Naoyuki*; Maeda, Shigetaka
Journal of Nuclear Science and Technology, 61(4), p.509 - 520, 2024/04
Ac is attracting attention as an alpha-emitting medical radioisotope. Since its demand is expected to increase, domestic production of Ac is required from the viewpoint of Japan's medical research and economic security. To establish the technical bases for the Ac production, JAEA has evaluated the radioactivity that can be produced in the experimental fast reactor Joyo and designed the concept that upgrades the existing facilities for transporting the irradiated target from Joyo to a neighboring PIE facility rapidly. Efficient Ac Separation from Ra irradiated in a fast reactor was studied. Ba and La were used as alternatives to Ra and Ac, respectively. By using DGA resin as an adsorbent, it can be expected that Ra and impurities generated by irradiation will be removed and Ac will be isolated. This study has revealed that Joyo can sufficiently produce Ac as a raw material for pharmaceuticals.
Konno, Chikara
JAEA-Conf 2023-001, p.143 - 146, 2024/02
I modified NJOY2016.67 to produce photonuclear ACE files which can be used in MCNP6.2 and PHITS3.27 and produced the ACE file of the JENDL-5 photonuclear sub-library. Simple test calculations with the produced ACE file supported that the produced ACE file had no serious problems.
Pyeon, C. H.*; Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro
Nuclear Science and Engineering, 197(11), p.2902 - 2919, 2023/11
Times Cited Count:1 Percentile:63.33(Nuclear Science & Technology)Sample reactivity and void reactivity experiments are carried out in the solid-moderated and solid-reflected cores at the Kyoto University Critical Assembly (KUCA) with the combined use of aluminum (Al), lead (Pb) and bismuth (Bi) samples, and Al spacers simulating the void. MCNP6.2 eigenvalue calculations together with JENDL-4.0 provide good accuracy of sample reactivity with the comparison of experimental results; also experimental void reactivity is attained by using MCNP6.2 together with JENDL-4.0 and ENDF/B-VII.1 with a marked accuracy of relative difference between experiments and calculations. Uncertainty quantification of sample reactivity and void reactivity is acquired by using the sensitivity coefficients based on MCNP6.2/ksen and covariance library data of SCALE6.2 together with ENDF/B-VII.1, arising from the impact of uncertainty induced by Al, Pb and Bi cross sections. A series of reactivity analyses with the Al spacer simulating the void demonstrates the means of analyzing the void in the solid-moderated and solid-reflected cores at KUCA
Konno, Chikara; Ota, Masayuki*; Kwon, Saerom*; Onishi, Seiki*; Yamano, Naoki*; Sato, Satoshi*
Journal of Nuclear Science and Technology, 60(9), p.1046 - 1069, 2023/09
Times Cited Count:5 Percentile:97.88(Nuclear Science & Technology)JENDL-5 was validated from a viewpoint of shielding applications under the Shielding Integral Test Working Group of the JENDL Committee. The following benchmark experiments were selected: JAEA/FNS in-situ experiments, Osaka Univ./OKTAVIAN TOF experiments, ORNL/JASPER sodium experiments, NIST iron experiment and QST/TIARA experiments. These experiments were analyzed with MCNP and nuclear data libraries (JENDL-5, JENDL-4.0 or JENDL-4.0/HE, ENDF/B-VIII.0 and JEFF-3.3). The analysis results demonstrate that JENDL-5 is comparable to or better than JENDL-4.0 or JENDL-4.0/HE, ENDF/B-VIII.0 and JEFF-3.3.
Kimura, Yoshiki; Tsuchiya, Kenichi*
Radioisotopes, 72(2), p.121 - 139, 2023/07
Rapid and precise radioisotope identification in the scene of nuclear detection and nuclear security incidents is one of the challenging issues for the prompt response on the detection alarm or the incidents. A radioisotope identification algorithm using a deep artificial neural network model applicable to handheld gamma-ray detectors has been proposed in the present paper. The proposed algorithm automatically identifies gamma-emitting radioisotopes based on the count contribution ratio (CCR) from each of them estimated by the deep artificial neural network model trained by simulated gamma-ray spectra. The automated radioisotope identification algorithm can support first responders of nuclear detection and nuclear security incidents without sufficient experience and knowledge in radiation measurement. The authors tested the performance of the proposed algorithm using two different types of deep artificial neural network models in application to handheld detectors having high or low energy resolution. The proposed algorithm showed high performance in identifying artificial radioisotopes for actually measured gamma-ray spectra. It was also confirmed that the algorithm is applicable to identifying U and automated uranium categorization by analyzing estimated CCRs by the deep artificial neural network models. The authors also com-pared the performance of the proposed algorithm with a conventional radioisotope identification method and discussed promising ways to improve the performance of the algorithm using the deep artificial neural network.
Ho, H. Q.; Ishii, Toshiaki; Nagasumi, Satoru; Ono, Masato; Shimazaki, Yosuke; Ishitsuka, Etsuo; Goto, Minoru; Simanullang, I. L.*; Fujimoto, Nozomu*; Iigaki, Kazuhiko
Nuclear Engineering and Design, 396, p.111913_1 - 111913_9, 2022/09
Times Cited Count:1 Percentile:27.23(Nuclear Science & Technology)Ho, H. Q.; Ishitsuka, Etsuo; Iigaki, Kazuhiko
Recent Contributions to Physics, 82(3), p.16 - 20, 2022/09
Nagata, Hiroshi; Otsuka, Kaoru; Omori, Takazumi; Ide, Hiroshi
JAEA-Technology 2022-017, 113 Pages, 2022/08
Japan Materials Testing Reactor (JMTR) was decided as a one of decommission facilities in April 2017. The activation activity of radioactive materials remaining in the reactor facility was evaluated in order to submit the decommissioning plan to the Nuclear Regulation Authority. Total activation activity was 9.310 Bq after the permanent shutdown of reactor, 2.710 Bq after 21 years, 1.010 Bq after 40 years and 2.410 Bq after 100 years. The structure with high activation activity was the core structural materials in JMTR such as beryllium frame, aluminum reflector, etc., and the material was stainless steel, beryllium, etc. The ratio of nuclides to the total amount of activated radioactivity was highest in H-3 until about 40 years after the reactor shutdown, and then in Ni-63. For reference, the radioactivity level was classified based on the results of the obtained radioactivity concentration. The ratio of the weight of each radioactivity level to the total weight was 0.3-0.4% (10-13t) for L1, 0.0-0.4% (0-14t) for L2, 1.0-1.2% (32-39t) for L3 and 98.0-98.7% (about 3200t) for CL until 100 years after the reactor shutdown. It was found that those classified as CL account for more than 90% of the total. When treating and disposing of radioactive waste, evaluation will be carried out based on appropriate methods, including evaluation results of secondary pollutants.
Kochiyama, Mami; Sakai, Akihiro
JAEA-Technology 2022-009, 56 Pages, 2022/06
It is necessary to evaluate radioactivity inventory in wastes before disposal of low-level radioactive wastes generated from dismantling research reactors. It is efficient for owners of each research reactor to use a common radioactive evaluation method in order to comply with the license application for disposal facility. In this report, neutron transport and activation calculations were carried out for the Rikkyo University research reactor in order to examine a common radioactivity evaluation method for burial disposal of radioactive wastes generated by dismantling. We adopted the neutron transport codes DORT and MCNP and the activation code ORIGEN-S with cross-section libraries based on JENDL-4.0 and JENDL/AD-2017. The radioactivity concentrations obtained by the radiochemical analysis and both calculation codes were in agreement by 0.4 to 3 times. Therefore, by appropriately considering this difference, the radioactivity evaluation method by DORT, MCNP and ORIGEN-S can be applied to the radioactivity evaluation for buried disposal. In order to classify wastes from dismantling by clearance or buried disposal method according to their radioactivity levels, we also created radioactivity concentration distributions in the concrete area and graphite thermal column area.
Fujimoto, Nozomu*; Tada, Kenichi; Ho, H. Q.; Hamamoto, Shimpei; Nagasumi, Satoru; Ishitsuka, Etsuo
Annals of Nuclear Energy, 158, p.108270_1 - 108270_8, 2021/08
Times Cited Count:3 Percentile:43.41(Nuclear Science & Technology)Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03
Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by Li(n,)H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.
Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01
Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a research and testing reactors containing beryllium reflectors.
Tsai, P.-E.; Heilbronn, L. H.*; Lai, B.-L.*; Iwata, Yoshiyuki*; Murakami, Takeshi*; Sheu, R.-J.*
Nuclear Instruments and Methods in Physics Research B, 449, p.62 - 70, 2019/06
Times Cited Count:3 Percentile:31.66(Instruments & Instrumentation)The secondary neutrons produced from 100- and 230-MeV/nucleon He ions, respectively, stopping in the thick iron, PMMA and water targets are measured by the time of flight method combined with the pulse shape discrimination of liquid scintillators. The experimental data show that the secondary neutrons were contributed from breakup of projectiles, emission from overlap regions of projectile nuclei and target nuclei, and evaporation of projectiles and target nuclei. The measured double-differential thick target neutron yields, angular distributions, and total neutron yields per ion, were benchmarked by the PHITS, FLUKA, and MCNP model calculations. The default models in these codes agree the experimental data well at intermediate-to-large angles in the low-to-intermediate energy range. However, the physics models implemented in PHITS need further improvement for some particular nuclear interaction mechanisms, and this work can be an importance reference for future model development.
Yeom, Y. S.*; Han, M. C.*; Choi, C.*; Han, H.*; Shin, B.*; Furuta, Takuya; Kim, C. H.*
Health Physics, 116(5), p.664 - 676, 2019/05
Times Cited Count:7 Percentile:60.64(Environmental Sciences)Recently, Task Group 103 of the ICRP developed the mesh-type reference computational phantoms (MCRPs), which are planned for use in future ICRP dose coefficient calculation. Performance of major Monte Carlo particle transport codes (Geant4, MCNP6, and PHITS) were tested with MCRP. External and internal exposure of various particles and energies were calculated and the computational times and required memories were compared. Additionally calculation for voxel-mesh phantom was also conducted so that the influence of different mesh-representation in each code was studied. Memory usage of MRCP was as large as 10 GB with Geant4 and MCNP6 while it is much less with PHITS (1.2 GB). In addition, the computational time required for MRCP tends to increase compared to voxel-mesh phantoms with Geant4 and MCNP6 while it is equal or tends to decrease with PHITS.
Kondo, Ryoichi*; Endo, Tomohiro*; Yamamoto, Akio*; Tada, Kenichi
Proceedings of International Conference on Mathematics and Computational Methods applied to Nuclear Science and Engineering (M&C 2019) (CD-ROM), p.1493 - 1502, 2019/00
A perturbation capability of ACE formatted cross section files was developed using the modules of FRENDY. Uncertainty quantification using MCNP was carried out for the Godiva critical experiment by the RS method. We verified the results of the RS method by comparing with those obtained by the conventional sensitivity analyses. Moreover, uncertainty reduction using the bias factor method with the RS technique was applied to kinetic parameter, i.e., neutron generation time.
Ishitsuka, Etsuo; Kenzhina, I.*; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
JAEA-Technology 2018-010, 33 Pages, 2018/11
As a part of study on the mechanism of tritium release to the primary coolant in research and testing reactors, tritium recoil release rate from Li and U impurities in the neutron reflector made by beryllium, aluminum and graphite were calculated by PHITS code. On the other hand, the tritium production from Li and U impurities in beryllium neutron reflectors for JMTR and JRR-3M were calculated by MCNP6 and ORIGEN2 code. By using both results, the amount of recoiled tritium from beryllium neutron reflectors were estimated. It is clear that the amount of recoiled tritium from Li and U impurities in beryllium neutron reflectors are negligible, and 2 and 5 orders smaller than that from beryllium itself, respectively.
Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*
EPJ Nuclear Sciences & Technologies (Internet), 4, p.42_1 - 42_7, 2018/11
This paper presents a nuclear data sensitivity and uncertainty analysis of the effective delayed neutron fraction for critical and subcritical cores of the MYRRHA reactor using the continuous-energy Monte Carlo transport code MCNP. The sensitivities are calculated by the modified -ratio method proposed by Chiba. Comparing the sensitivities obtained with different scaling factors introduced by Chiba shows that a value of is the most suitable for the uncertainty quantification of . Using the calculated sensitivities and the JENDL-4.0u covariance data, the uncertainties for the critical and subcritical cores are determined to be 2.2 0.2% and 2.0 0.2%, respectively, which are dominated by delayed neutron yield of Pu and U.
Iwamoto, Hiroki; Stankovskiy, A.*; Fiorito, L.*; Van den Eynde, G.*
Journal of Nuclear Science and Technology, 55(5), p.539 - 547, 2018/05
Times Cited Count:9 Percentile:66.49(Nuclear Science & Technology)The applicability of Monte Carlo techniques, namely the Monte Carlo sensitivity method and the random-sampling method, for uncertainty quantification of the effective delayed neutron fraction is investigated using the continuous-energy Monte Carlo transport code, MCNP, from the perspective of statistical convergence issues. This study focuses on the nuclear data as one of the major sources of uncertainty. For validation of the calculated , a critical configuration of the VENUS-F zero-power reactor was used. It is demonstrated that Chiba's modified -ratio method is superior to Bretscher's prompt -ratio method in terms of reducing the statistical uncertainty in calculating not only but also its sensitivities and the uncertainty due to nuclear data. From this result and a comparison of uncertainties obtained by the Monte Carlo sensitivity method and the random-sampling method, it is shown that the Monte Carlo sensitivity method using Chiba's modified -ratio method is the most practical for uncertainty quantification of . Finally, total uncertainty due to nuclear data for the VENUS-F critical configuration is determined to be approximately 2.7% with JENDL-4.0u, which is dominated by the delayed neutron yield of U.
Han, M. C.*; Yeom, Y. S.*; Lee, H. S.*; Shin, B.*; Kim, C. H.*; Furuta, Takuya
Physics in Medicine & Biology, 63(9), p.09NT02_1 - 09NT02_9, 2018/05
Times Cited Count:7 Percentile:38.85(Engineering, Biomedical)The multi-threading computation performances of the Geant4, MCNP6, and PHITS codes were evaluated using three tetrahedral-mesh phantoms with different complexity. Photon and neutron transport simulations were conducted and the initialization time, calculation time, and memory usage were measured as a function of the number of threads N used in the simulation. The initialization time significantly increases with the complexity of the phantom, but not much with the number of the threads. For the calculation time, Geant4 showed good parallelization efficiency with multi-thread computation (30 times speed-up factor for N = 40) adopting the private tallies while saturation of the speed-up factor were observed in MCNP6 and PHITS (10 and a few times for N = 40) due to the time delay for the sharing tallies. On the other hand, Geant4 requires larger memory specification and the memory usage rapidly increases with the number of threads compared to MCNP6 or PHITS. It is notable that when compared to the other codes, the memory usage of PHITS is much smaller, regardless of both the complexity of the phantom and the number of the threads.
Ho, H. Q.; Honda, Yuki; Goto, Minoru; Takada, Shoji
Annals of Nuclear Energy, 112, p.42 - 47, 2018/02
Times Cited Count:8 Percentile:61.87(Nuclear Science & Technology)