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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.

Komeda, Masao; Ozu, Akira; Mori, Takamasa; Nakatsuka, Yoshiaki; Maeda, Makoto; Kureta, Masatoshi; Toh, Yosuke

Journal of Nuclear Science and Technology, 55(8), P. 962, 2018/08

Times Cited Count：0 Percentile：100(Nuclear Science & Technology)We correct the derivation of equations in the derivation of equations in the paper of "Study of the neutron multiplication effect in an active neutron methods [J Nucl Sci Technol. 2017;54(11):1233-1239]". Although the derivations are not correct, the obtained equations are correct. Therefore, the results and discussions of the paper remain the same.

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：2 Percentile：50.89(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.

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

JAEA-Data/Code 2016-019, 450 Pages, 2017/03

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.

Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

JAEA-Data/Code 2016-018, 421 Pages, 2017/03

In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants, etc. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions.

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

Annals of Nuclear Energy, 82, p.85 - 89, 2015/08

Times Cited Count：6 Percentile：31.94(Nuclear Science & Technology)This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 52(2), p.282 - 293, 2015/02

Times Cited Count：0 Percentile：100(Nuclear Science & Technology)We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the "finite neutron multiplication factor", , deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty. The developed method is useful for validating the nuclear design methodology concerning void reactivity.

Yoshioka, Kenichi*; Kikuchi, Tsukasa*; Gunji, Satoshi*; Kumanomido, Hironori*; Mitsuhashi, Ishi*; Umano, Takuya*; Yamaoka, Mitsuaki*; Okajima, Shigeaki; Fukushima, Masahiro; Nagaya, Yasunobu; et al.

Journal of Nuclear Science and Technology, 50(6), p.606 - 614, 2013/06

Times Cited Count：1 Percentile：85.35(Nuclear Science & Technology)We have developed an intra-pellet neutron flux and conversion ratio distribution measurement method. A foil activation method with special foils was used for the neutron flux distribution measurement. A -ray spectrum analysis method with special collimators was used for the conversion ratio distribution measurement. Using the developed methods, intra-pellet neutron flux distributions and conversion ratio distributions were measured in critical experiments on a reduced-moderation LWR. Measured values were analyzed with a deterministic method and a Monte Carlo method. The neutron flux distribution measurements and analyses agreed within the range of 1% to 2%. The conversion ratio distribution measurements and analyses were consistent with each other. We found that the measurement methods are useful for the validation of neutron behavior in a fuel pellet, which is known as micro reactor physics.

Okajima, Shigeaki; Kugo, Teruhiko; Mori, Takamasa

Genshiryoku Kyokasho "Genshiro Butsurigaku", 258 Pages, 2012/03

This textbook is prepared for a lecture of the reactor physics in Nuclear Professional School of The University of Tokyo. In Chapter 1, the characteristics of nucleus and the interaction between nucleus and neutron were described. In Chapter 2, the fission reaction, its chain reaction and a concept on criticality were described. Chapter 3 described about the fundamental study on a diffusion equation to treat the neutron spatial distributions in a substance. In Chapter 4, the application of a diffusion equation was shown to learn the feature of neutron distribution in a substance with homogeneous composition. Chapter 5 described the physics of neutron slowing-down in a substance and Chapter 6 investigated the neutron slowing-downed to become in thermal equilibrium state in the substance. Chapter 7 considered the effect of the practical heterogeneous composition from the homogeneous one to the criticality. The JENDL-4.0 library was referred for the nuclear reaction cross section data.

Nagaya, Yasunobu; Mori, Takamasa

Progress in Nuclear Science and Technology (Internet), 2, p.842 - 850, 2011/10

Applicability of the Taylor series approach with the arbitrary-order differential operator sampling (DOS) method is examined for the calculation of sample reactivity worth. The DOS method is extended to obtain the differential coefficients of the effective multiplication factor and the perturbed source effect up to the arbitrary order. The methodology is implemented into a continuous-energy Monte Carlo code MVP to perform benchmark calculations. It is found that the second-order Taylor series approach gives an enough accurate result for simple fast systems of Godiva and Jezebel. A discrepancy of 10% is, on the other hand, observed for the Np sample worth calculation for the tank-type critical assembly TCA even with the fifth-order Taylor series approach. The perturbed source effect has a significant contribution for the calculation of sample worth reactivity; it must be estimated for all the cases. Alternative approaches are also examined for the TCA sample worth problem. They give similar results as the ordinary higher-order Taylor series approach.

Sakurai, Takeshi; Kosako, Kazuaki*; Mori, Takamasa

Progress in Nuclear Science and Technology (Internet), 2, p.318 - 329, 2011/10

Chiba, Go; Nagaya, Yasunobu; Mori, Takamasa

Journal of Nuclear Science and Technology, 48(8), p.1163 - 1169, 2011/08

Times Cited Count：3 Percentile：67.22(Nuclear Science & Technology)The effective delayed neutron fraction can be accurately calculated with the continuous-energy Monte Carlo method using the iterated fission probability (IFP) if the sufficiently large number of generations is considered. In order to deterministically quantify the required number of generations in the IFP-based calculations, the concept of the generation-dependent importance functions is introduced to calculations. Furthermore, the most appropriate reactor property used in the IFP calculations, which reduces the required number of generations, is theoretically derived. Through numerical calculations, it is shown that the several generations are required in the IFP-based calculations and that the use of the appropriate reactor property can reduce the required number of generations. An efficient procedure for the IFP-based calculations with the Monte Carlo method is also proposed.

Sakurai, Takeshi; Mori, Takamasa; Suzaki, Takenori*; Okajima, Shigeaki; Ando, Yoshihira*; Yamamoto, Toru*; Liem, P. H.*

Journal of Nuclear Science and Technology, 48(5), p.816 - 825, 2011/05

Times Cited Count：2 Percentile：76.59(Nuclear Science & Technology)Mori, Takamasa; Nakajima, Norihiro

Nippon Genshiryoku Gakkai-Shi, 53(3), P. 227, 2011/03

no abstracts in English

Nagaya, Yasunobu; Mori, Takamasa

Annals of Nuclear Energy, 38(2-3), p.254 - 260, 2011/02

Times Cited Count：21 Percentile：10.28(Nuclear Science & Technology)Alternative methods are proposed to estimate the effective delayed neutron fraction in Monte Carlo calculations: the eigenvalue methods jointly used with the differential operator sampling and correlated sampling techniques. In particular, the eigenvalue method with the differential operator sampling technique has a distinct feature that it gives a theoretically exact value. To verify the proposed methods, Monte Carlo calculations are performed for several systems with simple geometry. It is found that the results obtained with the proposed methods agree with the reference deterministic results within sufficiently small statistical uncertainties. The perturbed source effect must be taken into account to estimate an exact value.

Nagaya, Yasunobu; Chiba, Go; Mori, Takamasa; Irwanto, D.*; Nakajima, Ken*

Annals of Nuclear Energy, 37(10), p.1308 - 1315, 2010/10

Times Cited Count：20 Percentile：14.19(Nuclear Science & Technology)Monte Carlo calculation methods to estimate the effective delayed neutron fraction are investigated: One is proposed by Meulekamp et al. and the other is by Nauchi et al. It is revealed that both the methods calculate the delayed neutron fraction weighted with the importance functions defined by Kobayashi. The accuracy of the methods are also examined for several simple benchmark systems. Consequently, it is found that Meulekamp's method causes 5% discrepancies in the values for fast systems; Nauchi's method gives good results for fast bare systems but 10% discrepancies for fast reflected systems. Both the methods calculate the values approximately within the accuracy of 2% for thermal systems.

Nagaya, Yasunobu; Mori, Takamasa

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 6 Pages, 2010/10

Nuclear reactor analysis requires calculations of reactivity worth such as control rod worth, void reactivity worth, sample reactivity worth, etc. It is, however, difficult to perform such calculations with the Monte Carlo method if the worth is small. In the present work, we extend the differential operator sampling method such that the derivative terms up to the fourth-order and the perturbed source effect up to the fourth-order can be estimated, and examine the applicability of the fourth-order differential operator sampling method to reactivity worth calculations. As a benchmark calculation, we perform the reactivity worth calculation for Godiva. The perturbation is introduced by decreasing the density in the central region of a radius of 1 cm. The result with the fourth-order differential operator sampling method agrees well with the reference one. As a practical calculation, we also perform Np sample worth calculation for TCA. The calculated sample reactivity worth with the differential sampling method approaches to the reference value as the higher-order effect is taken into account up to the fourth order. However, there still exists a discrepancy of 17%. It is, thus, found that many more higher-order effects must be taken into account or another method must be applied in this case.

Sakurai, Takeshi; Kosako, Kazuaki*; Mori, Takamasa

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 8 Pages, 2010/10

Mori, Takamasa; Nagaya, Yasunobu

Journal of Nuclear Science and Technology, 46(8), p.793 - 798, 2009/08

In order to investigate the impact of resonance elastic scattering models on the Doppler reactivity effect, the exact and the constant cross section models with the thermal motion of target nucleus were newly implemented into the MVP-2 continuous-energy Monte Carlo code with and without consideration of energy-dependent resonance cross section, respectively, and the UO pin-cell Doppler reactivity benchmark calculations were carried out with the modified code and the JENDL-3.3 library. The present study has revealed that the exact model gives more negative Doppler reactivity coefficients by 711% than the conventional asymptotic model, while the constant cross section model gives slightly less negative coefficients than the conventional asymptotic model. Furthermore, it is found that the impact of resonance elastic scattering models is considerably large around the resonances where elastic scattering has relatively high contribution, whereas capture-dominant resonances have no significant impacts on the Doppler reactivity coefficient.

Sakurai, Takeshi; Mori, Takamasa; Okajima, Shigeaki; Tani, Kazuhiro*; Suzaki, Takenori*; Saito, Masaki*

Journal of Nuclear Science and Technology, 46(6), p.624 - 640, 2009/06

Times Cited Count：5 Percentile：57.79(Nuclear Science & Technology)