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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

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 (183 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

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：83.53(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

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

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：1 Percentile：42.23(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：1 Percentile：24.18(Nuclear Science & Technology)Mori, Takamasa; Kojima, Kensuke*; Suyama, Kenya

JAEA-Research 2018-010, 57 Pages, 2019/02

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：73(Chemistry, Inorganic & Nuclear)Fukaya, Yuji; Goto, Minoru; Nishihara, Tetsuo

Nuclear Engineering and Design, 326, p.108 - 113, 2018/01

Times Cited Count：3 Percentile：47(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：5 Percentile：59.32(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：0 Percentile：0.08JAEA 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：20 Percentile：95.35(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

Nagaya, Yasunobu; Adachi, Masaaki*

Proceedings of International Conference on Mathematics & Computational Methods Applied to Nuclear Science & Engineering (M&C 2017) (USB Flash Drive), 6 Pages, 2017/04

MVP is a general-purpose Monte Carlo code for neutron and photon transport calculations based on the continuous-energy method. To speed up the MVP code, hybrid parallelization is applied with a message passing interface library MPI and a shared-memory multiprocessing library OpenMP. The performance test has been done for an eigenvalue calculation of a fast reactor subassembly, a fixed-source calculation of a neutron/photon coupled problem and a PWR full core model. Comparisons has been made for MPI only with 4 processes and hybrid parallelism with 4 processes 3 threads. As a result, the hybrid parallelism yields the reduction of elapsed time by 16% to 34% and the used memories are almost the same.

Kuwagaki, Kazuki*; Nagaya, Yasunobu

JAEA-Data/Code 2017-007, 27 Pages, 2017/03

The integral benchmark test of JENDL-4.0 for U-233 systems using the continuous-energy Monte Carlo code MVP was conducted. The previous benchmark test was performed only for U-233 thermal solution and fast metallic systems in the ICSBEP handbook. In this study, MVP input files were prepared for uninvestigated benchmark problems in the handbook including compound thermal systems (mainly lattice systems) and integral benchmark test was performed. The prediction accuracy of JENDL-4.0 was evaluated for effective multiplication factors ('s) of the U-233 systems. As a result, a trend of underestimation was observed for all the categories of U-233 systems. In the benchmark test of ENDF/B-VII.1 for U-233 systems with the ICSBEP handbook, it is reported that a decreasing trend of calculated values in association with a parameter ATFF (Above-Thermal Fission Fraction) is observed. The ATFF values were also calculated in this benchmark test of JENDL-4.0 and the same trend as ENDF/B-VII.1 was observed.

Ohgama, Kazuya; Nakano, Yoshihiro; Oki, Shigeo

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

Times Cited Count：1 Percentile：14.48(Nuclear Science & Technology)The power distribution and core characteristics in various configurations of fuel subassemblies with an innerduct structure in the Japan Sodium-cooled Fast Reactor were evaluated using a Monte Carlo code for neutron transport and burnup calculation. The correlation between the fraction of fuel subassemblies facing outward and the degree of power increase at the core center was observed regardless of the compositions. This indicated that the spatial fissile distribution caused by innerduct configurations was the major factor of the difference in the power distribution. A power increase was also found in an off-center region, and it tended to be greater than that at the core center because of the steep gradient of neutron flux intensity. The differences in the worth of control rods caused by innerduct configurations were confirmed.

Yamamoto, Kento*; Akie, Hiroshi; Suyama, Kenya; Hosoyamada, Ryuji*

JAEA-Technology 2015-019, 110 Pages, 2015/10

In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. The recent development of higher-enrichment fuel has enhanced the benefit of the application of Burnup Credit. In the present study, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study.