Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Yanagisawa, Hiroshi; Umeda, Miki; Motome, Yuiko; Murao, Hiroyuki
JAEA-Technology 2022-030, 80 Pages, 2023/02
JAEA-Technology-2022-030.pdf:2.57MB
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.
Cs activity in spent nuclear fuel with calculated detector response functionSato, Shunsuke*; Nauchi, Yasushi*; Hayakawa, Takehito*; Kimura, Yasuhiko; Kashima, Takao*; Futakami, Kazuhiro*; Suyama, Kenya
Journal of Nuclear Science and Technology, 9 Pages, 2022/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)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.
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.
Yanagisawa, Hiroshi
JAEA-Technology 2021-023, 190 Pages, 2021/11
JAEA-Technology-2021-023.pdf:5.25MB
Computational analyses on nuclear criticality characteristics were carried out for heterogeneous lattice systems composed of water moderator and fuel rods utilized in low-power research and test reactors, in which the depletion of fuel due to burnup is relatively small, by using the continuous-energy Monte Carlo code MVP Version 2 with the evaluated nuclear data library JENDL-4.0. In the analyses, the minimum critical number of fuel rods was evaluated using calculated neutron multiplication factors for the heterogeneous systems of the uranium dioxide fuel rod in the Static Experiments Critical Facility (STACY) and the Tank-type Critical Assembly (TCA), and the uranium-zirconium hydride fuel rod in the Nuclear Safety Research Reactor (NSRR). In addition, six sorts of the ratio of reaction rates, which are components of neutron multiplication factors, were calculated in the analyses to explain the variation of neutron multiplication factors with the ratio of water moderator to fuel volume in a unit fuel rod cell. Those results of analyses are considered to be useful for the confirmation of reasonableness and validity of criticality safety measures as data showing criticality characteristics for water-moderated heterogeneous lattice systems composed of the existing fuel rods in research and test reactors, of which criticality data are not sufficiently provided by the Criticality Safety Handbook.
Ishitsuka, Etsuo; Mitsui, Wataru*; Yamamoto, Yudai*; Nakagawa, Kyoichi*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Nagasumi, Satoru; Takamatsu, Kuniyoshi; Kenzhina, I.*; et al.
JAEA-Technology 2021-016, 16 Pages, 2021/09
JAEA-Technology-2021-016.pdf:1.8MB
As a summer holiday practical training 2020, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the downsizing of reactor core were studied by the MVP-BURN. As a result, it is clear that a 1.6 m radius reactor core, containing 54 (18
3 layers) fuel blocks with 20% enrichment of 
U, and BeO neutron reflector, could operate continuously for 30 years with thermal power of 5 MW. Number of fuel blocks of this compact core is 36% of the HTTR core. As a next step, the further downsizing of core by changing materials of the fuel block will be studied.
Ikeda, Reiji*; Ho, H. Q.; Nagasumi, Satoru; Ishii, Toshiaki; Hamamoto, Shimpei; Nakano, Yumi*; Ishitsuka, Etsuo; Fujimoto, Nozomu*
JAEA-Technology 2021-015, 32 Pages, 2021/09
JAEA-Technology-2021-015.pdf:2.74MB
Burnup calculation of the HTTR considering temperature distribution and detailed burning regions was carried out using MVP-BURN code. The results show that the difference in k, as well as the difference in average density of some main isotopes, is insignificant between the cases of uniform temperature and detailed temperature distribution. However, the difference in local density is noticeable, being 6% and 8% for U and 
Pu, respectively, and even 30% for the burnable poison 
B. Regarding the division of burning regions to more detail, the change of k is also small of 0.6%
k/k or less. The small burning region gives a detailed distribution of isotopes such as U, Pu, and B. As a result, the effect of graphite reflector and the burnup behavior could be evaluated more clearly compared with the previous study.
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：1 Percentile：30.57(Nuclear Science & Technology)Ishitsuka, Etsuo; Nakashima, Koki*; Nakagawa, Naoki*; Ho, H. Q.; Ishii, Toshiaki; Hamamoto, Shimpei; Takamatsu, Kuniyoshi; Kenzhina, I.*; Chikhray, Y.*; Matsuura, Hideaki*; et al.
JAEA-Technology 2020-008, 16 Pages, 2020/08
JAEA-Technology-2020-008.pdf:2.98MB
As a summer holiday practical training 2019, the feasibility study for nuclear design of a nuclear battery using HTTR core was carried out, and the U enrichment and burnable poison of the fuel, which enables continuous operation for 30 years with thermal power of 5 MW, were studied by the MVP-BURN. As a result, it is clear that a fuel with U enrichment of 12%, radius of burnable poison and natural boron concentration of 1.5 cm and 2wt% are required. As a next step, the downsizing of core will be studied.
Nagasumi, Satoru; Matsunaka, Kazuaki*; Fujimoto, Nozomu*; Ishii, Toshiaki; Ishitsuka, Etsuo
JAEA-Technology 2020-003, 13 Pages, 2020/05
JAEA-Technology-2020-003.pdf:1.5MB
The influence of the control rod model on the nuclear characteristics of the HTTR has been evaluated, by creating detailed control rod model, in which geometric shape was close to that of the actual control rod structure, in MVP code. According to refinement of the control rod model, the critical control rod position was 11 mm lower than that of the conventional model, and this was close to the measured value of 1775 mm. The reactivity absorbed by the shock absorber located at the tip of the control rod was 0.2%k/k, and this was 14 mm difference at the critical control rod position. Considering the effect of refinement of the control rod and the effect of the shock absorber, the correction amount for the analysis value in SRAC code due to the shape effect of the control rod, is -0.05%k/k in reactivity, and -3 mm in the critical control rod position at low temperature criticality.
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
Komeda, Masao; Toh, Yosuke
Annals of Nuclear Energy, 135, p.106993_1 - 106993_6, 2020/01
Times Cited Count：2 Percentile：33.86(Nuclear Science & Technology)This paper presents a conceptual study of a novel active method using a neutron source. The main feature of this new method is the fast rotation of a neutron source in order to derive the fission neutron counts and applying the counts to detect the nuclear material. Irradiating neutrons to a container that involves nuclear material, the measurement data include both neutrons from the neutron source and fission neutrons. However, if the neutron source is rotated quite fast, the components of the irradiation neutrons and fission neutrons are separated. Since this novel method does not require an expensive D-T tube, this new system is expected to be affordable and easy to assemble.
Maeda, Makoto; Furutaka, Kazuyoshi; Kureta, Masatoshi; Ozu, Akira; Komeda, Masao; Toh, Yosuke
Journal of Nuclear Science and Technology, 56(7), p.617 - 628, 2019/07
Times Cited Count：2 Percentile：27.54(Nuclear Science & Technology)Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya
JAEA-Research 2018-010, 57 Pages, 2019/02
JAEA-Research-2018-010.pdf:6.25MB
In order to estimate applicability of the statistical geometry model (STGM) of MVP/GMVP, a parametric study in infinite geometry and criticality safety analyses for direct disposal of spent fuel in simple finite geometry have been carried out by using the MVP Monte Carlo code. It has been found that calculations with STGM for larger fuel spheres give larger thermal utilization factors and larger infinite multiplication factors compared with explicit random models in the range of fuel sphere packing fraction between 6.5 % and 63.3 %. Substantial differences are not observed between the results with two nearest neighbor distributions (NNDs); that given by the MCRDF code and the analytical expression based on a statistically uniform distribution. It is inferred that the overestimation by STGM is caused by the facts that STGM cannot take account of the surroundings of each neutron, whether a fuel sphere rich region or a water moderator rich one, because STGM always uses an NND averaged over such surroundings and that STGM, therefore, cannot take the effect of consecutive scatterings in the water moderator into account.
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
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：6 Percentile：59.55(Chemistry, Inorganic & Nuclear)Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo
Nuclear Engineering and Design, 326, p.108 - 113, 2018/01
Times Cited Count：3 Percentile：35.39(Nuclear Science & Technology)Burn-up characteristics and criticality of impurity contained into graphite structure for commercial scale prismatic High Temperature Gas-cooled Reactor (HTGR) have been investigated. For HTGR, of which the core is filled graphite structure, the impurity contained into the graphite has unignorable poison effect for criticality. Then, GTHTR300, commercial scale HTGR, employed high grade graphite material named IG-110 to take into account the criticality effect for the reflector blocks next to fuel blocks. The fuel blocks, which should also employ IG-110, employ lower grade graphite material named IG-11 from the economic perspective. In this study, the necessity of high grade graphite material for commercial scale HTGR is reconsidered by evaluating the burn-up characteristics and criticality of the impurity. The poison effect of the impurity, which is used to be expressed by a boron equivalent, reduces exponentially like burn-up of B, and saturate at a level of 1 % of the initial value of boron equivalent. On the other hand, the criticality effect of the boron equivalent of 0.03 ppm, which corresponds to a level of 1 % of IG-11 shows ignorable values lower than 0.01 %k/kk' for both of fuel blocks and reflector blocks. The impurity can be represented by natural boron without problem. Therefore, the poison effect of the impurity is evaluated with whole core burn-up calculations. As a result, it is concluded that the impurity is not problematic from the viewpoint of criticality for commercial scale HTGR because it is burned clearly until End of Cycle (EOC) even with the low grade graphite material of IG-11. According to this result, more economic electricity generation with HTGR is expected by abolishing the utilization of IG-110.
Komeda, Masao; Ozu, Akira; Mori, Takamasa; Nakatsuka, Yoshiaki; Maeda, Makoto; Kureta, Masatoshi; Toh, Yosuke
Journal of Nuclear Science and Technology, 54(11), p.1233 - 1239, 2017/11
Times Cited Count：7 Percentile：62.15(Nuclear Science & Technology)The previous active neutron method cannot remove the influence of the multiplication effect of neutrons produced by second- and subsequent fission reactions, and it might overestimate the amount of nuclear material if an item contains large amounts. In this paper, we discussed the correction method for the neutron multiplication effect on the measured data in the fast neutron direct interrogation (FNDI) method, one of the active neutron methods, supposing that the neutron multiplication effect is caused mainly by third-generation neutrons from the second-fission reactions under the condition that the forth-generation neutrons are much fewer. This paper proposed a correction method for the neutron multiplication effect in the measured data. Moreover we have shown a possibility that this correction method gives rough estimates of the effective neutron multiplication factor and the subcriticality.
Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio
EPJ Web of Conferences, 146, p.02028_1 - 02028_5, 2017/09
Times Cited Count：3 Percentile：88.62JAEA has started to develop new nuclear data processing system FRENDY (FRom Evaluated Nuclear Data libralY to any application). In this presentation, the outline of the development of FRENDY is presented. And functions and performances of FRENDY are demonstrated by generation and validation of the continuous energy cross section data libraries for MVP, PHITS and MCNP codes.
Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio
Journal of Nuclear Science and Technology, 54(7), p.806 - 817, 2017/07
Times Cited Count：35 Percentile：97.71(Nuclear Science & Technology)JAEA has developed an evaluated nuclear data library JENDL and several nuclear analysis codes such as MARBLE2, SRAC, MVP and PHITS. Though JENDL and these computer codes have been widely used in many countries, the nuclear data processing system to generate the data library for application programs had not been developed in Japan and foreign nuclear data processing systems, e.g., NJOY and PREPRO are used. To process the new library for JAEA's computer codes immediately and independently, JAEA started to develop the new nuclear data processing system FRENDY in 2013. In this paper, outline, function, and verification of FRENDY are described.
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
