Refine your search:     
Report No.
Search Results: Records 1-20 displayed on this page of 239

Presentation/Publication Type

Initialising ...


Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...


Initialising ...


Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Japanese Evaluated Nuclear Data Library version 5; JENDL-5

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.

Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01

Journal Articles

Report on the 34th Meeting of Working Party on International Nuclear Data Evaluation Co-operation (WPEC)

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Tada, Kenichi; Sugawara, Takanori; Yokoyama, Kenji

Kaku Deta Nyusu (Internet), (133), p.1 - 6, 2022/10

no abstracts in English

Journal Articles

Development of ARKADIA for the innovation of advanced nuclear reactor design process (Overview of optimization process development in design optimization support tool, ARKADIA-Design)

Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07

To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.

Journal Articles

Completion of JENDL-5 and prospects for its application to numerical analysis, 4; Integral test of JENDL-5; Benchmark analysis in fast reactor system

Yokoyama, Kenji; Taninaka, Hiroshi

Kaku Deta Nyusu (Internet), (132), p.25 - 33, 2022/06

This article explains the results of integral test of JENDL-5 by benchmark analysis in fast reactor system, which were presented in a special session of the 2022 Spring Annual Meeting of the Atomic Energy Society of Japan (AESJ). The latest version of Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. In order to confirm the applicability of JENDL-5 to the fast reactor system, we conducted a set of benchmark analysis using the integral experiment data included in the fast reactor nuclear design database which is being developed by JAEA. With respect to major nuclear characteristics in the standard fast reactor system, it was confirmed that the ratios of analysis result and experimental result (C/E values) based on JENDL-5 were almost the same as those of JENDL-4.0. In the special session, the results of sensitivity analysis were reported. Since the results have been described in the proceedings of the AESJ meeting, we add the results of the versions under development of JENDL-5 and discuss their relationship with the reported results of sensitivity analysis.

Journal Articles

Applicability study of Bayesian optimization in core neutronic design using a toy model

Kuwagaki, Kazuki; Yokoyama, Kenji

Proceedings of International Conference on Physics of Reactors 2022 (PHYSOR 2022) (Internet), 10 Pages, 2022/05

In Japan Atomic Energy Agency (JAEA), an innovative design approach named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) for the advanced nuclear reactors is currently under development. One of the tasks in ARKADIA is to build a system that automatically optimizes core and fuel designs by conducting core neutronic and thermal-hydraulic calculations, fuel integrity evaluations, and plant dynamic analyses. This system will be implemented to automatically find an optimal design that minimizes (or maximizes) objective function defined by a core performance, while varying the core and fuel design parameters such as fuel pin diameter, core height and diameter. In this study, as the first step of the system development, we focused only on core neutronic design and conducted a study of automatic optimization. As the optimization algorithm, Bayesian optimization (BO), which is an effective method for optimization problems with expensive computation cost of objective function, was selected. The applicability of BO was studied based on single- and two-objective optimization examples of core neutronic design in a toy model. As a result, in the former, it was confirmed that BO can obtain the optimal solution, which well matches the reference solution calculated by a brute force calculation, with a small number of required calculation executions. Its usability on core neutronic designs, where the computation cost per case is large, was confirmed. In the latter, it was shown that BO can obtain a pareto solutions-set that shows good agreement with the reference solution.

JAEA Reports

Development of the unified cross-section set ADJ2017R

Yokoyama, Kenji; Maruyama, Shuhei; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2021-019, 115 Pages, 2022/03


In JAEA, several versions of unified cross-section set for fast reactors have been developed so far; we have developed a new unified cross-section set ADJ2017R, which is an improved version of the unified cross-section setADJ2017 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 in reactor core design via the cross-section adjustment methodology; the values are stored in the standard database for FBR core design. In the methodology, the cross-section set is adjusted by integrating the information such as uncertainty (covariance) of nuclear data, uncertainty of integral experiment / analysis, sensitivity of integral experiment with respect to nuclear data. ADJ2017R basically has the same performance as ADJ2017, but we conducted an additional investigation on ADJ2017 and revised the following two points. The first is to unify the evaluation method of the correlation coefficient of uncertainty caused by experiments (hereinafter referred to as the experimental correlation coefficient). Because it was found that the common uncertainty used in the evaluation of the experimental correlation coefficient was evaluated by two different methods, the experimental correlation coefficients were revised for all experimental data, and the evaluation method was unified. The second is the review of the integral experiment data used for the cross-section adjustment calculation. It was found that one of the experimental values of composition ratio after irradiation of the Am-243 sample has a problem in uncertainty evaluation because its experimental uncertainty is extremely small compared to the others. The cross-section adjustment calculation was, therefore, redone by excluding the experimental value. In the creation of ADJ2017, a total of 719 data sets were analyzed and evaluated, and eventually adopted 620 integral experimental data sets. In contrast, a total of 61

Journal Articles

Development of numerical analysis codes for multi-level and multi-physics approaches in an advanced reactor design study

Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03

Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.

Journal Articles

Development of evaluation method for core deformation reactivity feedback in sodium-cooled fast reactor by coupled analysis approach

Doda, Norihiro; Uwaba, Tomoyuki; Yokoyama, Kenji; Nemoto, Toshiyuki*; Tanaka, Masaaki

Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 14 Pages, 2022/03

In sodium-cooled fast reactors, reactivity feedback is generated by thermal deformation of the core fuel assembly during core temperature rise. To utilize the core deformation reactivity as an inherent safety characteristic and to eliminate excessive conservativeness of core design in the safety evaluation, an evaluation method by coupling analyses of neutronics, thermal-hydraulics, and structural deformation has been developed. An experiment of unprotected loss-of-flow event in the experimental fast breeder reactor EBR-II was analyzed. The analysis results show that the core deformation reactivity has a negative feedback effect, and that the deformation reactivity is affected not only by the fuel movement but also by the movement of reflectors around the fuel. As a result, the availability of the evaluation method for core deformation reactivity feedback by coupled analysis approach is confirmed.

Journal Articles

Great achievements of M. Salvatores for nuclear data adjustment study with use of integral experiments

Yokoyama, Kenji; Ishikawa, Makoto*

Annals of Nuclear Energy, 154, p.108100_1 - 108100_11, 2021/05

 Times Cited Count:1 Percentile:30.57(Nuclear Science & Technology)

In the design of innovative nuclear reactors such as fast reactors, the improvement of the prediction accuracies for neutronics properties is an important task. The nuclear data adjustment is a promising methodology for this issue. The idea of the nuclear data adjustment was first proposed in 1964. Toward its practical application, however, a great deal of study has been conducted over a long time. While it took about 10 years to establish the theoretical formulation, the research and development for its practical application has been conducted for more than half a century. Researches in this field are still active, and the fact suggests that the improvement of the prediction accuracies is indispensable for the development of new types of nuclear reactors. Massimo Salvatores, who passed away in March 2020, was one of the first proposers to develop the nuclear data adjustment technique, as well as one of the great contributors to its practical application. Reviewing his long-time works in this area is almost the same as reviewing the history of the nuclear data adjustment methodology. The authors intend that this review would suggest what should be done in the future toward the next development in this area. The present review consists of two parts: a) the establishment of the nuclear data adjustment methodology and b) the achievements related to practical applications. Furthermore, the former is divided into two aspects: the study on the nuclear data adjustment theory and the numerical solution for sensitivity coefficient that is requisite for the nuclear data adjustment. The latter is separated to three categories: the use of integral experimental data, the uncertainty quantification and design target accuracy evaluation, and the promotion of nuclear data covariance development.

Journal Articles

Development of neutronics, thermal-hydraulics, and structure mechanics coupled analysis method on integrated numerical analysis for design optimization support in fast reactor

Doda, Norihiro; Uwaba, Tomoyuki; Nemoto, Toshiyuki*; Yokoyama, Kenji; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 26, 4 Pages, 2021/05

For design optimization of fast reactors, in order to consider the feedback reactivity due to thermal deformation of the core when the core temperature rises, which could not be considered in the conventional design analysis, a neutronics, thermal-hydraulics, and structure mechanics coupled analysis method has been developed. Neutronics code, plant dynamics code, and structural mechanics code are coupled by a control module in python script. This paper outlines the coupling method of analysis codes and the results of its application to an experiment in an actual plant.

JAEA Reports

Development of burnup/depletion calculation code based on ORIGEN2 cross-section libraries and Chebyshev rational approximation method, CRAMO

Yokoyama, Kenji; Jin, Tomoyuki*

JAEA-Data/Code 2021-001, 47 Pages, 2021/03


A new burnup/depletion calculation code, CRAMO, was developed by combining an ORIGEN2 cross-section library set, ORLIB, based on Japanese evaluated nuclear data library, JENDL, and a burnup/depletion solver based on Chebyshev rational approximation method. CRAMO uses the ORIGEN2 cross-section library set ORLIBJ40 based on JENDL-4.0, and the burnup/depletion solver implemented in the versatile reactor analysis code system, MARBLE. It was confirmed that results of CRAMO agreed well with those of ORIGEN2 for burnup/depletion and radioactivity calculation cases. The development of CRAMO made it possible to use ORLIB without using ORIGEN2. It will be possible to provide an easy-to-use processed JENDL data set for burnup/depletion and radioactivity calculations in combination with a burnup/depletion based on Chebyshev rational approximation method. The present version of CRAMO is a subset of ORIGEN2 and can compute only compositions and radioactivities after irradiation. However, since various kinds of outputs of ORIGEN2 can be evaluated by using the composition, it is possible to reproduce many functions of ORIGEN2 by adding post-processing modules.

JAEA Reports

Annual report on the environmental radiation monitoring around the Tokai Reprocessing Plant FY2019

Nakano, Masanao; Fujii, Tomoko; Nemoto, Masashi; Tobita, Keiji; Seya, Natsumi; Nishimura, Shusaku; Hosomi, Kenji; Nagaoka, Mika; Yokoyama, Hiroya; Matsubara, Natsumi; et al.

JAEA-Review 2020-069, 163 Pages, 2021/02


Environmental radiation monitoring around the Tokai Reprocessing Plant has been performed by the Nuclear Fuel Cycle Engineering Laboratories, based on "Safety Regulations for the Reprocessing Plant of Japan Atomic Energy Agency, Chapter IV - Environmental Monitoring". This annual report presents the results of the environmental monitoring and the dose estimation to the hypothetical inhabitant due to the radioactivity discharged from the plant to the atmosphere and the sea during April 2019 to March 2020. In this report, some data include the influence of the accidental release from the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Co., Inc. (the trade name was changed to Tokyo Electric Power Company Holdings, Inc. on April 1, 2016) in March 2011. Appendices present comprehensive information, such as monitoring programs, monitoring methods, monitoring results and their trends, meteorological data and discharged radioactive wastes. In addition, the data which were influenced by the accidental release and exceeded the normal range of fluctuation in the monitoring, were evaluated.

Journal Articles

Benchmarks of depletion and decay heat calculation between MENDEL and MARBLE

Yokoyama, Kenji; Lahaye, S.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.109 - 116, 2020/10

CEA/DEN/DM2S/SERMA and JAEA/NSEC are working on benchmarks for burnup, isotopic concentrations and decay heat calculations in the collaboration framework between both organisms. Both actors of this benchmark are independently developing their own simulation code systems for computing quantities of interest in nuclear fuel cycle domain: MENDEL in CEA and MARBLE in JAEA. The purpose of the benchmark is to verify each system by comparing both calculation results on specific applications. MENDEL uses a several solvers for the resolution of Bateman equation. Runge-Kutta method or Chebyshev Rational Approximation method (CRAM) are used for irradiation computations. An analytical solver can also be used for decay calculations. MARBLE can use Krylov subspace method or CRAM method. As the first phase of the benchmark, we compared the calculated results of decay heat and isotropic concentrations following by a Pu-239 fast fission pulse. We applied nuclear data from three libraries: (1) JEFF-3.1.1, (2) JENDL/DDF-2015 + JENDL/FPY-2011, and (3) ENDF/B-VII.1. Nuclear data and burnup chain were generated from these libraries independently on each system. We confirmed that the results for both systems were in very good agreement with each other. Numerical results were also compared to experimental data. As the second phase of the benchmark, we are proceeding with a burnup calculation benchmark of MENDEL and MARBLE using the nuclear data and burnup chain provided by ORLIBJ33, which is a set of cross-section data based on JENDL-3.3 for ORIGEN-2 code system. We will also compare with calculation results by the ORIGEN-2 code with ORLIBJ33. Since the series of ORLIB, that is, ORLIBJ32, ORLIBJ33, and ORLIBJ40, have been widely used especially in Japan for many years, the comparison with ORLIB is effective for confirming the performance of MENDEL and MARBLE.

Journal Articles

HPRL; International cooperation to identify and monitor priority nuclear data needs for nuclear applications

Dupont, E.*; Bossant, M.*; Capote, R.*; Carlson, A. D.*; Danon, Y.*; Fleming, M.*; Ge, Z.*; Harada, Hideo; Iwamoto, Osamu; Iwamoto, Nobuyuki; et al.

EPJ Web of Conferences, 239, p.15005_1 - 15005_4, 2020/09

 Times Cited Count:3 Percentile:96.13

Journal Articles

Overview of the OECD-NEA Working Party on International Nuclear Data Evaluation Cooperation (WPEC)

Fleming, M.*; Bernard, D.*; Brown, D.*; Chadwick, M. B.*; De Saint Jean, C.*; Dupont, E.*; Ge, Z.*; Harada, Hideo; Hawari, A.*; Herman, M.*; et al.

EPJ Web of Conferences, 239, p.15002_1 - 15002_4, 2020/09

 Times Cited Count:0 Percentile:0.1

Journal Articles

An Electron-capture efficiency in femtosecond filamentation

Nakashima, Nobuaki*; Yatsuhashi, Tomoyuki*; Sakota, Kenji*; Iwakura, Izumi*; Hashimoto, Sena*; Yokoyama, Keiichi; Matsuda, Shohei

Chemical Physics Letters, 752, p.137570_1 - 137570_5, 2020/08

Photo-redox reactions between Eu$$^{3+}$$ and Eu$$^{2+}$$ ions are induced by laser irradiation in alcoholic solution. Efficiency, wavelength dependence, and laser-power dependence are investigated with three different lasers. Nano second laser pulses at a wavelength of 308 nm is found to cause one-photon redox reactions with a quantum yield around 0.5. Nano second laser pulses at a wavelength of 394 nm induces two-photon reduction of Eu$$^{3+}$$ to form Eu$$^{2+}$$. When the pulse energy is 5 mJ, the quantum yield is measured to be 0.015. Although the quantum yield is one order of magnitude lower than that of the one photon reduction, reduction phenomena can be easily observed under the moderate laser field strength. Because of the two-photon nature, there should be a room to improve the efficiency by increasing the laser field strength.

Journal Articles

The Joint evaluated fission and fusion nuclear data library, JEFF-3.3

Plompen, A. J. M.*; Cabellos, O.*; De Saint Jean, C.*; Fleming, M.*; Algora, A.*; Angelone, M.*; Archier, P.*; Bauge, E.*; Bersillon, O.*; Blokhin, A.*; et al.

European Physical Journal A, 56(7), p.181_1 - 181_108, 2020/07

 Times Cited Count:180 Percentile:99.41(Physics, Nuclear)

The Joint Evaluated Fission and Fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides $$^{235}$$U, $$^{238}$$U and $$^{239}$$Pu, on $$^{241}$$Am and $$^{23}$$Na, $$^{59}$$Ni, Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yileds, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 is excellent for a wide range of nuclear technology applications, in particular nuclear energy.

Journal Articles

Development of neutronics and thermal-hydraulics coupled analysis method on platform for design optimization in fast reactor

Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki

Keisan Kogaku Koenkai Rombunshu (CD-ROM), 25, 4 Pages, 2020/06

With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.

Journal Articles

Report of 31st Meeting of the Working Party on International Nuclear Data Evaluation Co-operation (WPEC)

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kimura, Atsushi; Yokoyama, Kenji; Tada, Kenichi

Kaku Deta Nyusu (Internet), (124), p.23 - 34, 2019/10

The 31st annual meeting and the subgroup meeting of the Working Party on International Nuclear Data Evaluation Co-operation (WPEC) under the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) was held at the head quarter of OECD/NEA located at Boulogne-Billancourt near Paris from 24 to 28 in June in 2019. The activities about nuclear data measurement and evaluation of each region or country were reported at the annual meeting, and the SG activities were discussed at the subgroup meetings. The summary of these meetings are reported.

Journal Articles

Burnup calculation with versatile reactor analysis code system MARBLE2 (interactive execution demo)

Yokoyama, Kenji

Nihon Genshiryoku Gakkai Dai-51-Kai Robutsuri Kaki Semina Tekisuto "Nensho Keisan No Kiso To Jissen", p.95 - 135, 2019/08

The burnup calculation function included in the versatile reactor analysis code system system MARBLE2 is introduced by an interactive execution demo. Although the main purpose of MARBLE2 is to analyze nuclear characteristics of fast reactors, the users can use it while assembling small functions according to purpose. Therefore, it can be applied other purposes than the nuclear characteristic analysis of fast reactors. In order to realize such usage, MARBLE is developed by using an object-oriented scripting language Python. As the Python implementation is short and easy to understand, the burnup function of MARBLE is explained by showing several examples of the implementation. In addition, an example of constructing a simple burnup calculation system using MARBLE is introduced.

239 (Records 1-20 displayed on this page)