Watanabe, Shoto*; Minato, Futoshi; Kimura, Masaaki*; Iwamoto, Nobuyuki
Journal of Nuclear Science and Technology, 59(11), p.1399 - 1406, 2022/11
Journal of Nuclear Science and Technology, 59(10), p.1232 - 1241, 2022/10
Nakayama, Hiromasa; Onodera, Naoyuki; Satoh, Daiki; Nagai, Haruyasu; Hasegawa, Yuta; Idomura, Yasuhiro
Journal of Nuclear Science and Technology, 59(10), p.1314 - 1329, 2022/10
We developed a local-scale high-resolution atmospheric dispersion and dose assessment system (LHADDAS) for safety and consequence assessment of nuclear facilities and emergency response to nuclear accidents or deliberate releases of radioactive materials in built-up urban areas. This system is composed of pre-processing of input files, main calculation by local-scale high-resolution atmospheric dispersion model using large-eddy simulation (LOHDIM-LES) and real-time urban dispersion simulation model based on a lattice Boltzmann method (CityLBM), and post-processing of dose-calculation by simulation code powered by lattice dose-response functions (SIBYL). LHADDAS has a broad utility and offers superior performance in (1) simulating turbulent flows, plume dispersion, and dry deposition under realistic meteorological conditions, (2) performing real-time tracer dispersion simulations using a locally mesh-refined lattice Boltzmann method, and (3) estimating air dose rates of radionuclides from air concentrations and surface deposition in consideration of the influence of individual buildings and structures. This system is promising for safety assessment of nuclear facilities as an alternative to wind tunnel experiments, detailed pre/post-analyses of a local-scale radioactive plume dispersion in case of nuclear accidents, and quick response to emergency situations resulting from deliberate release of radioactive materials by a terrorist attack in an urban central district area.
Sato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Kashima, Takao*; Futakami, Kazuhiro*; Suyama, Kenya
Journal of Nuclear Science and Technology, 9 Pages, 2022/10
A new non-destructive method for evaluating Cs activity in spent nuclear fuels was proposed and experimentally demonstrated for physical measurements in burnup credit implementation. Cs activities were quantified using gamma ray measurements and numerical detector response simulations without reference fuels, in which 137Cs activities are well known. Fuel samples were obtained from a lead use assembly (LUA) irradiated in a commercial pressurized water reactor (PWR) up to 53 GWd/t. Gamma rays emitted from the samples were measured using a bismuth germinate (BGO) scintillation detector through a collimator attached to a hot cell. The detection efficiency of gamma rays with the detector was calculated using the PHITS particle transport calculation code considering the measurement geometry. The relative activities of Cs, Cs, and Eu in the sample were measured with a high-purity germanium (HPGe) detector for more accurate simulations of the detector response for the samples. The absolute efficiency of the detector was calibrated by measuring a standard gamma ray source in another geometry. Cs activity in the fuel samples was quantified using the measured count rate and detection efficiency. The quantified Cs activities agreed well with those estimated using the MVP-BURN depletion calculation code.
Ariyoshi, Gen; Obayashi, Hironari; Sasa, Toshinobu
Journal of Nuclear Science and Technology, 59(9), p.1071 - 1088, 2022/09
Electromagnetic induction method is one of the effective techniques for local velocity measurement in heavy liquid metals. Ricou and Vives' probe and Von Weissenfluh's probe are famous instrumentations using a permanent magnet. However, sensitivity and measurement volume of the probes show unexpected variation since demagnetization of the magnet is occurred by temperature increase up to the Curie temperature. In this study, electromagnetic probe incorporating a miniature electromagnet was newly developed to overcome such unexpected variation. The diameter and the length of the sensor was 6 mm and 155 mm, respectively. The sensitivity and the measurement volume of the probe were assessed by measurement of local velocity of flowing mercury in a square channel. To clarify the validity for the measured velocity profiles, numerical velocity profiles were calculated and compared with experiment. And the validity for the measured velocity profiles were confirmed by calculated result.
Kondo, Toshiki; Toda, Taro*; Takeuchi, Junichi*; Kargl, F.*; Kikuchi, Shin; Muta, Hiroaki*; Oishi, Yuji*
Journal of Nuclear Science and Technology, 59(9), p.1139 - 1148, 2022/09
no abstracts in English
Motegi, Kosuke; Shibamoto, Yasuteru; Kukita, Yutaka
Journal of Nuclear Science and Technology, 59(8), p.1037 - 1046, 2022/08
Kumagai, Yuta; Kusaka, Ryoji; Nakada, Masami; Watanabe, Masayuki; Akiyama, Daisuke*; Kirishima, Akira*; Sato, Nobuaki*; Sasaki, Takayuki*
Journal of Nuclear Science and Technology, 59(8), p.961 - 971, 2022/08
We investigated potential degradation of fuel debris caused by HO, which is the oxidant of major impact from water radiolysis. We performed leaching experiments on different kinds of simulated debris comprising U, Fe, Cr, Ni, and Zr in an aqueous HO solution. Chemical analysis of the leaching solution showed that U dissolution was induced by HO. Raman analysis after the leaching revealed that uranyl peroxides were formed on the surface of the simulated debris. These results demonstrate that uranyl peroxides are possible alteration products of fuel debris from HO reaction. However, the sample in which the main uranium-containing phase was a U-Zr oxide solid solution showed much less uranium dissolution and no Raman signal of uranyl peroxides. Comparison of these results indicates that formation of an oxide solid solution of Zr with UO improves the stability of fuel debris against HO reaction.
Satoh, Daiki; Sato, Tatsuhiko
Journal of Nuclear Science and Technology, 59(8), p.1047 - 1060, 2022/08
In this research, we simulated the neutron-response functions and detection efficiencies of a liquid organic scintillator using the particle and heavy-ion transport code system (PHITS). We incorporated the algorithm and database of the neutron-response simulation code SCINFUL-QMD into PHITS. Then, we updated the total, elastic, and inelastic cross-section data of the hydrogen and carbon nuclei for neutrons and developed a new scorer to analyze the light outputs from a scintillator. The calculation results of the neutron-response functions and the detection efficiencies were compared with results of SCINFUL-QMD, the previous PHITS with the new scorer, and the reported measurements. It was found that the improved PHITS successfully reproduced the results calculated by SCINFUL-QMD, except for around 150 MeV where a discontinuity of detection-efficiency curve was observed in the SCINFUL-QMD values. Our results showed better agreement with the measured data than the results of the previous PHITS. The uncertainties of the detection efficiencies calculated by PHITS using the present extensions were estimated to be approximately 15% for neutrons in the energy region below 100 MeV.
Kaburagi, Masaaki; Shimazoe, Kenji*; Kato, Masahiro*; Kurosawa, Tadahiro*; Takahashi, Hiroyuki*
Journal of Nuclear Science and Technology, 59(8), p.983 - 992, 2022/08
Cantarel, V.; Yamagishi, Isao
Journal of Nuclear Science and Technology, 59(7), p.888 - 897, 2022/07
Nakamura, Shoji; Toh, Yosuke; Kimura, Atsushi; Hatsukawa, Yuichi*; Harada, Hideo
Journal of Nuclear Science and Technology, 59(7), p.851 - 865, 2022/07
The present study performed integral experiments of I using a fast-neutron source reactor "YAYOI" of the University of Tokyo to validate evaluated nuclear data libraries. The iodine-129 sample and flux monitors were irradiated by fast neutrons in the Glory hole of the YAYOI reactor. Reaction rates of I were obtained by measurement of decay gamma-rays emitted from I. The validity of the fast-neutron flux spectrum in the Glory hole was confirmed by the ratios of the reaction rates of flux monitors. The experimental reaction rate of I was compared with that calculated with both the fast-neutron flux spectrum and evaluated nuclear data libraries. The present study revealed that the evaluated nuclear data of I cited in JENDL-4.0 should be reduced as much as 18% in neutron energies ranging from 10 keV to 3 MeV, and supported the reported data by Noguere below 100 keV.
Hirata, Yuho; Sato, Tatsuhiko; Watanabe, Kenichi*; Ogawa, Tatsuhiko; Parisi, A.*; Uritani, Akira*
Journal of Nuclear Science and Technology, 59(7), p.915 - 924, 2022/07
The reliability of dose assessment with radiation detectors is an important feature in various fields, such as radiotherapy, radiation protection, and high-energy physics. However, many detectors irradiated by high linear energy transfer (LET) radiations exhibit decreased efficiency called the quenching effect. This quenching effect depends not only on the particle LET but strongly on the ion species and its microscopic pattern of energy deposition. Recently, a computational method for estimating the relative efficiency of luminescence detectors was proposed following analysis of microdosimetric specific energy distributions simulated using the particle and heavy ion transport code system (PHITS). This study applied the model to estimate the relative optically stimulated luminescence (OSL) efficiency of BaFBr:Eu detectors. Additionally, we measured the luminescence intensity of BaFBr:Eu detectors exposed to He, C and Ne ions to verify the calculated data. The model reproduced the experimental data in the cases of adopting a microdosimetric target diameter of approximately 30-50 nm. The calculated relative efficiency exhibit ion-species dependence in addition to LET. This result shows that the microdosimetric calculation from specific energy is a successful method for accurately understanding the results of OSL measurements with BaFBr:Eu detectors irradiated by various particles.
Journal of Nuclear Science and Technology, 59(6), p.768 - 780, 2022/06
Sato, Yuki; Terasaka, Yuta
Journal of Nuclear Science and Technology, 59(6), p.677 - 687, 2022/06
Ohgama, Kazuya; Hara, Toshiharu*; Ota, Hirokazu*; Naganuma, Masayuki; Oki, Shigeo; Iizuka, Masatoshi*
Journal of Nuclear Science and Technology, 59(6), p.735 - 747, 2022/06
Chiba, Go*; Yamamoto, Akio*; Tada, Kenichi
Journal of Nuclear Science and Technology, 8 Pages, 2022/06
A new multi-group neutronics analysis sequence ACE-FRENDY-CBZ is proposed. This sequence is free from uses of any application libraries; with the ACE files as the starting point, multi-group cross section data of media comprising a target system are calculated with the FRENDY code, and multi-group neutron transport calculations are performed with modules of the CBZ code system. The ACE-FRENDY-CBZ sequence was tested against the eight fast neutron systems, and good agreement with the reference Monte Carlo results was obtained within 30 pcm differences in the bare systems and the thorium-reflected system, and approximately 100 pcm differences in the uranium-reflected systems. The use of the current-weighted total cross sections in the multi-group neutron transport calculations had non-negligible impacts over 100 pcm on k-eff, and the calculations with the current-weighted total cross sections systematically underestimated k-eff in the uranium-reflected systems.
Manabe, Kentaro; Sato, Kaoru; Takahashi, Fumiaki
Journal of Nuclear Science and Technology, 59(5), p.656 - 664, 2022/05
It is known that internal doses depend on the physical characteristics of an evaluation subject. Internal dose coefficients provided by the International Commission on Radiological Protection (ICRP) are evaluated using the characteristics of the standard Caucasian. It is important to grasp the variations of doses due to the differences in characteristics between Japanese and Caucasian when the dose coefficients of ICRP are applied to Japanese. This study evaluated dose coefficients using specific absorbed fraction (SAF) data based on the average adult Japanese physique which was developed by modification of the existing Japanese SAF data with additional calculations to make the existing data fit to the current dosimetric methodology of ICRP and compared them to those provided by ICRP. As a result, the discrepancies in dose coefficients were smaller than plus or minus 10% in most intake conditions. However, some intake conditions indicated varieties over 40% due to the differences in organ masses, amount of adipose tissues around the thoracic cavity, and so on. This information is useful in application of ICRP's dose coefficients to population of which physical characteristics are different from those of Caucasian. Further, the Japanese SAF data is published as an appendix of this paper.
Iwamoto, Yosuke; Hashimoto, Shintaro; Sato, Tatsuhiko; Matsuda, Norihiro; Kunieda, Satoshi; elik, Y.*; Furutachi, Naoya*; Niita, Koji*
Journal of Nuclear Science and Technology, 59(5), p.665 - 675, 2022/05
A benchmark study of PHITS3.24 has been conducted using neutron-shielding experiments listed in the Shielding Integral Benchmark Archive and Database. Five neutron sources were selected, which are generated from (1) 43- and 68-MeV proton-induced reaction on a thin lithium target, (2) 52-MeV proton-induced reaction on a thick graphite target, (3) 590-MeV proton-induced reaction on a thick lead target, (4) 500-MeV proton-induced reaction on a thick tungsten target, and (5) 800-MeV proton-induced reaction on a thick tantalum target. For all cases, overall agreements in the results are satisfactory when using the JENDL-4.0/HE to simulate neutron- and proton-induced reactions up to 200 MeV. However, discrepancies using PHITS default settings are observed in the results. For an accurate neutron-shielding design for accelerator facilities, using JENDL-4.0/HE in the particle and heavy-ion transport code system calculation is favorable.
Rovira Leveroni, G.; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Iwamoto, Osamu; Iwamoto, Nobuyuki; Katabuchi, Tatsuya*; Kodama, Yu*; Nakano, Hideto*; Sato, Yaoki*; et al.
Journal of Nuclear Science and Technology, 59(5), p.647 - 655, 2022/05