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Nagaya, Yasunobu; Ueki, Taro; Tonoike, Kotaro

Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09

A new Monte Carlo solver Solomon has been developed for the application to fuel-debris systems. It is designed not only for usual criticality safety analysis but also for criticality calculations of damaged reactor core including fuel debris. This paper describes the current status of Solomon and demonstrates the applications of the randomized Weierstrass function (RWF) model and the RWF model superimposed voxel geometry.

Tonoike, Kotaro; Watanabe, Tomoaki; Gunji, Satoshi; Yamane, Yuichi; Nagaya, Yasunobu; Umeda, Miki; Izawa, Kazuhiko; Ogawa, Kazuhiko

Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 9 Pages, 2019/09

Criticality control of the fuel debris in the Fukushima Daiichi Nuclear Power Station would be a risk-informed control to mitigate consequences of criticality events, instead of a deterministic control to prevent such events. The Nuclear Regulation Authority of Japan has administrated a research and development program to tackle this challenge since 2014. The Nuclear Safety Research Center of Japan Atomic Energy Agency, commissioned by the authority, is conducting activities such as computations of criticality characteristics of the fuel debris, development of a criticality analysis code, preparation of criticality experiments, and development of a criticality risk analysis method.

Yokoyama, Kenji; Sugino, Kazuteru; Ishikawa, Makoto; Maruyama, Shuhei; Nagaya, Yasunobu; Numata, Kazuyuki*; Jin, Tomoyuki*

JAEA-Research 2018-011, 556 Pages, 2019/03

We have developed a new unified cross-section set ADJ2017, which is an improved version of the unified cross-section set ADJ2010 for fast reactors. The unified cross-section set is used for reflecting information of C/E values (analysis / experiment values) obtained by integral experiment analyses; the values are stored in the standard database for FBR core design via the cross-section adjustment methodology, which integrates with the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. The ADJ2017 is based on Japan's latest nuclear data library JENDL-4.0 as in the previous version of ADJ2010, and it incorporates more information on integral experimental data sets related to minor actinides (MAs) and degraded plutonium (Pu). In the creation of ADJ2010, a total of 643 integral experimental data sets were analyzed and evaluated, and 488 of the integral experimental data sets were finally selected to be used for the cross-section adjustment. In contrast, we have evaluated a total of 719 data sets, and eventually adopted 620 integral experimental data sets to create ADJ2017. ADJ2017 shows almost the same performance as ADJ2010 for the main neutronic characteristics of conventional sodium-cooled MOX-fuel fast reactors. In addition, for the neutronic characteristics related to MA and degraded Pu, ADJ2017 improves the C/E values of the integral experimental data sets, and reduces the uncertainty induced by the nuclear data. ADJ2017 is expected to be widely used in the analysis and design research of fast reactors. Moreover, it is expected that the integral experimental data sets used for ADJ2017 can be utilized as a standard database of FBR core design.

Tada, Kenichi; Kunieda, Satoshi; Nagaya, Yasunobu

JAEA-Data/Code 2018-014, 106 Pages, 2019/01

A new nuclear data processing code FRENDY has been developed in order to process the evaluated nuclear data library JENDL. Development of FRENDY helps to disseminate JENDL and various nuclear calculation codes. FRENDY is developed not only to process the evaluated nuclear data file but also to implement the FRENDY functions to other calculation codes. Users can easily use many functions e.g., read, write, and process the evaluated nuclear data file, in their own codes when they implement the classes of FRENDY to their codes. FRENDY is coded with considering maintainability, modularity, portability and flexibility. The processing method of FRENDY is similar to that of NJOY. The current version of FRENDY treats the ENDF-6 format and generates the ACE file which is used for the continuous energy Monte Carlo codes such as PHITS and MCNP. This report describes the nuclear data processing methods and input instructions for FRENDY.

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：100JAEA 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：9 Percentile：19.56(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.

Nagaya, Yasunobu

Hoshasen, 43(2), p.49 - 54, 2017/05

no abstracts in English

Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Maruyama, Shuhei; Yokoyama, Kenji; Sugino, Kazuteru; Nagaya, Yasunobu; Oki, Shigeo

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 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.

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.

Fukushima, Masahiro; Kitamura, Yasunori*; Yokoyama, Kenji; Iwamoto, Osamu; Nagaya, Yasunobu; Leal, L. C.*

Proceedings of International Conference on the Physics of Reactors; Unifying Theory and Experiments in the 21st Century (PHYSOR 2016) (USB Flash Drive), p.605 - 619, 2016/05

A nuclear data of U has been recently evaluated for the CIELO (Collaborative International Evaluated Library Organization) project. We tested the newly-evaluated data of U using integral experiments of the Fast Critical Assembly (FCA) performed at JAEA. We selected two integral data of uranium-fueled FCA assemblies; one is the sodium-void reactivity worth of FCA XXVII-1 assembly and the other is the criticalities of the seven assemblies of FCA IX. The benchmark tests support the evaluation done in the resonance regions. However, the U capture cross section above the unresolved resonance range needs further investigation.

Brun, E.*; Zoia, A.*; Trama, J. C.*; Lahaye, S.*; Nagaya, Yasunobu

Proceedings of International Conference on Nuclear Criticality Safety (ICNC 2015) (DVD-ROM), p.351 - 360, 2015/09

This paper presents a joint work conducted at CEA Saclay and JAEA Tokai aimed at comparing the Monte Carlo codes TRIPOLI and MVP on a selection of ICSBEP benchmarks. Our goal is to establish a common set of Monte Carlo input decks, as a basis for rigorous inter-code comparison in criticality-safety. As a reference, we will use the MCNP Criticality Validation Suite: other Monte Carlo developers might easily join that effort in the future. For the purpose of inter-code comparison, the TRIPOLI and MVP input decks have been translated from those of MCNP, without any further assumptions. Both TRIPOLI and MVP have been run with the same ENDF/B-VII.0 evaluated nuclear data, and as far as possible the same simulation options as in the original LANL work. In this abstract, we present preliminary results on the BIGTEN benchmark. In the full paper these will be extended to the 31 benchmarks of the MCNP Criticality Validation Suite. In the future, this database will also help in the analysis of sensitivity to nuclear data, CPU times and figures of merit.

Nagaya, Yasunobu

Annals of Nuclear Energy, 82, p.226 - 229, 2015/08

Times Cited Count：4 Percentile：53.79(Nuclear Science & Technology)The methods to calculate the kinetics parameters of and with the differential operator sampling have been reviewed. The comparison of the results obtained with the differential operator sampling and iterated fission probability approaches has been performed. It is shown that the differential operator sampling approach gives the same results as the iterated fission probability approach within the statistical uncertainty. In addition, the prediction accuracy of the evaluated nuclear data library JENDL-4.0 for the measured values is also examined. It is shown that JENDL-4.0 gives a good prediction except for the uranium-233 systems. The present results imply the need for revisiting the uranium-233 nuclear data evaluation but detailed sensitivity analysis is required for further discussion.

Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

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

Times Cited Count：6 Percentile：38.87(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*; Kitada, Takanori*; Nagaya, Yasunobu

Progress in Nuclear Energy, 82, p.7 - 15, 2015/07

Times Cited Count：1 Percentile：86.11(Nuclear Science & Technology)A reduced-moderation LWR has been developed for the reduction of spent fuel and for the efficient utilization of uranium resources. The streaming channel concept to improve the negative void reactivity coefficient is one of the features of the reactor. This concept makes the fuel assembly more heterogeneous. The geometrical heterogeneity makes azimuthal neutron flux distribution of fuel rods steep. To validate azimuthal neutron flux distribution calculation, we measured the distribution around fuel rods in reduced moderation LWR lattices. These measurements were conducted in NCA with the foil activation method. The core consisted of the central triangular tight lattice zone and the outer driver zone of a square lattice. A pile of polystyrene plates for simulating void fraction was installed into the triangular tight lattice. Azimuthal neutron flux distributions were deduced from the activity of these small foils measured with plastic scintillators. Measurements were compared to calculations by the MVP code with JENDL-3.3. It was found that calculations agreed well with measurements.

Nagaya, Yasunobu

Proceedings of Joint International Conference on Mathematics and Computation, Supercomputing in Nuclear Applications and the Monte Carlo Method (M&C + SNA + MC 2015) (CD-ROM), 9 Pages, 2015/04

A new method to calculate the prompt neutron decay constant () with the Monte Carlo method is proposed. It is based on the conventional - search algorithm but no iteration is required for the value search. The eigenvalue is expressed in the truncated Taylor series with regard to ; the differential coefficients are calculated with the differential operator sampling, which is one of the Monte Carlo perturbation techniques. In order to examine the applicability of the proposed method, verification has been performed for simple geometries of a bare fast system (Godiva) and an unreflected thermal system (STACY). Comparisons has been done with the pulsed neutron source (PNS) simulation and the direct calculation from the definition of the value. The results with the proposed method show good agreement with the reference PNS simulation.

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.

Nagaya, Yasunobu

Proceedings of International Conference on the Physics of Reactors; The Role of Reactor Physics toward a Sustainable Future (PHYSOR 2014) (CD-ROM), 13 Pages, 2014/09

Monte Carlo analysis has been performed to investigate the impact of the exact resonance elastic scattering model on the Doppler reactivity coefficient for the UO pin cell geometry with the parabolic temperature profile. As a result, the exact scattering model affects the coefficient similarly for both the flat and parabolic temperature profiles; it increases the contribution of uranium-238 (U) resonance capture in the energy region from 16 eV to 150 eV and does uniformly in the radial direction. Then the following conclusions hold for both the exact and asymptotic resonance scattering models. The Doppler reactivity coefficient is well reproduced with the definition of the effective fuel temperature (equivalent flat temperature) proposed by Grandi et al.