Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 64

Presentation/Publication Type

Initialising ...

Refine

Journal/Book Title

Initialising ...

Meeting title

Initialising ...

First Author

Initialising ...

Keyword

Initialising ...

Language

Initialising ...

Publication Year

Initialising ...

Held year of conference

Initialising ...

Save select records

Journal Articles

Evaluation of fixed absorber reactivity measurement in the prototype fast reactor Monju

Ohgama, Kazuya; Katagiri, Hiroki; Takegoshi, Atsushi*; Hazama, Taira

Nuclear Technology, 207(12), p.1810 - 1820, 2021/12

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

Numerical study on an interface compression method for the volume of fluid approach

Okagaki, Yuria; Yonomoto, Taisuke; Ishigaki, Masahiro; Hirose, Yoshiyasu

Fluids (Internet), 6(2), p.80_1 - 80_17, 2021/02

Journal Articles

Voltage drop analysis and leakage suppression design for mineral-insulated cables

Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Otsuka, Noriaki; Tsuchiya, Kunihiko

Journal of Nuclear Science and Technology, 57(12), p.1276 - 1286, 2020/12

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

The influence of materials of mineral-insulated (MI) cables on their electrical characteristics upon exposure to high-temperature conditions was examined via a transmission test, in the objective of achieving the stability of the potential distribution along the cable length. Occurrence of a voltage drop along the cable was confirmed for aluminum oxide (Al$$_{2}$$O$$_{3}$$) and magnesium oxide (MgO), as insulating materials of the MI cable. A finite-element method (FEM)-based analysis was performed to evaluate the leakage in the potentials, which was found at the terminal end. Voltage drop yields by the transmission test and the analysis were in good agreement for the MI cable of Al$$_{2}$$O$$_{3}$$ and MgO materials, which suggests the reproducibility of the magnitude relationship of the experimental results via the FEM analysis. To suppress the voltage drop, the same FEM analysis was conducted, the diameter of the core wires ($$d$$) and the distance between them ($$l$$) were varied. Considering the variation of $$d$$, the potential distribution in the MI cable produced a minimum voltage drop corresponding to a ratio $$d/D$$ of 0.35, obtained by dividing $$d$$ with that of the insulating material ($$D$$). In case of varying $$l$$, a minimum voltage drop was l/$$D$$ of 0.5.

Journal Articles

Assessment of nuclear simulation credibility

Tanaka, Masaaki; Nakada, Kotaro*; Kudo, Yoshiro*

Nihon Kikai Gakkai-Shi, 123(1222), p.26 - 29, 2020/09

In the nuclear engineering, simulations are used in radiation, thermal hydraulic, chemical reaction, and structural fields, and the integrated fields thereof, to be applied to the design, construction and operation of nuclear facilities. This article describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements described in the "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)". And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.

Journal Articles

Establishment of guideline for credibility assessment of nuclear simulations in the Atomic Energy Society of Japan

Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08

Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.

Journal Articles

A Validation study of a neutronics design methodology for fast reactors using reaction rate distribution measurements in the prototype fast reactor Monju

Ohgama, Kazuya; Takegoshi, Atsushi; Katagiri, Hiroki*; Hazama, Taira

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

Journal Articles

Validation and verification for the melting and eutectic models in JUPITER code

Chai, P.; Yamashita, Susumu; Nagae, Yuji; Kurata, Masaki

Proceedings of 9th Conference on Severe Accident Research (ERMSAR 2019) (Internet), 14 Pages, 2019/03

In order to obtain a precise understanding of molten material behavior inside RPV and to improve the accuracy of the SA code, a new computational fluid dynamics (CFD) code with multi-phase, multi-physics models, which is called JUPITER, was developed. It optimized the algorithms of the multi-phase calculation. Besides, the chemical reactions are also modeled carefully in the code so that the melting process could be treated precisely. A series of verification and validation studies are conducted, which show good agreement with analytical solutions and previous experiments. The capabilities of the multi-physics models in JUPITER code provide us another useful tool to investigate the molten material behaviors in the relevant severe accident scenario.

Journal Articles

Some characteristics of gas-liquid two-phase flow in vertical large-diameter channels

Shen, X.*; Schlegel, J. P.*; Hibiki, Takashi*; Nakamura, Hideo

Nuclear Engineering and Design, 333, p.87 - 98, 2018/07

 Times Cited Count:5 Percentile:30.29(Nuclear Science & Technology)

Journal Articles

State-of-the-art approach and issue to establish simulation credibility

Nakada, Kotaro*; Kudo, Yoshiro*; Koshizuka, Seiichi*; Tanaka, Masaaki

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(3), p.173 - 177, 2018/03

The Atomic Energy Society of Japan (AESJ) published "Guideline for Credibility Assessment of Nuclear Simulations 2015" in June, 2016 which specifies the concepts on methodology for the prediction with uncertainty quantification and the quality management based on the concept of verification and validation (V&V) of modeling and simulation. In this report, the outlines of activities in AESJ for publication of the guideline and the expectation for effective implementation of the guideline are described including that of the lectures with major respondents of the questionnaires.

Journal Articles

Simulation of fuel-coolant interaction SERENA2 test based on JASMINE version 3

Hotta, Akitoshi*; Morita, Akinobu*; Kajimoto, Mitsuhiro*; Maruyama, Yu

Nihon Genshiryoku Gakkai Wabun Rombunshi, 16(3), p.139 - 152, 2017/09

Journal Articles

Model verification and validation procedure for a neutronics design methodology of next generation fast reactors

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

Journal Articles

Report of lecture course for "Guideline for Credibility Assessment of Nuclear Simulations 2015"

Tanaka, Masaaki

Nihon Genshiryoku Gakkai Keisan Kagaku Gijutsu Bukai Nyusu Reta (Internet), (27), p.9 - 15, 2017/03

In this report, the outline of the contents in the lecture course for "Guideline for Credibility Assessment of Nuclear Simulations 2015" published in June, 2016 from the Atomic Energy Society of Japan (AESJ) and the result of the lectures held in Tokyo and Osaka are introduced with the results of the questionnaires from the audience.

Journal Articles

Activity to establish the guideline for credibility assessment of nuclear simulations in the Atomic Energy Society of Japan

Tanaka, Masaaki

Nihon Genshiryoku Gakkai Keisan Kagaku Gijutsu Bukai Nyusu Reta (Internet), (24), p.16 - 28, 2015/09

In order to enhance the simulation credibility, it is necessary to establish the procedure on verification and validation including the estimation of the modeling uncertainty. Lessons learned from the Fukushima Daiichi NPP Accident have indicated that it was important to recognize the credibility of the simulation. By putting forward to standardize the procedure on verification and validation including the estimation of the modeling uncertainty, it is expected to establish the basis of the simulation technology to realize the world highest level of nuclear safety and continuous improvement. The recent activity in the Atomic Energy Society of Japan (AESJ) for the guideline establishment is introduced.

Journal Articles

Inter-code comparison of TRIPOLI${textregistered}$ and MVP on the MCNP criticality validation suite

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${textregistered}$ 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${textregistered}$ and MVP input decks have been translated from those of MCNP, without any further assumptions. Both TRIPOLI${textregistered}$ 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.

Journal Articles

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.8141 - 8155, 2015/08

In this paper, the authors focus on four kinds of thermal-hydraulic issues associated with the SDC, i.e. fuel assembly thermal-hydraulics, natural circulation decay heat removal, thermal striping phenomena, and core disruptive accidents, and provide a description of their evaluation method developments including verification and validation and necessary experimental studies for the Japan Sodium-cooled Fast Reactor (JSFR). These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all phenomena envisioned in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing down of knowledge/technologies.

Journal Articles

Proposal of benchmark problem of thermal striping phenomena in planar triple parallel jets tests for fundamental code validation in sodium-cooled fast reactor development

Kobayashi, Jun; Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki; Kamide, Hideki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.6664 - 6677, 2015/08

Numerical simulation is recognized an essential tool for the physical phenomena analysis and plant design study of a sodium-cooled fast reactor (SFR). In order to enhance credibility of the numerical results in the activities for plant design by using numerical simulations, it is recognized that verification and validation (V&V) process is very important. In this study, experiments for planar triple parallel jets mixing phenomena conducted in JAEA were proposed as benchmark problems for the code validation in the area of thermal striping study in the SFR development.

JAEA Reports

Verification and validation of the THYTAN code for the graphite oxidation analysis in the HTGR systems

Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi

JAEA-Technology 2014-038, 51 Pages, 2014/12

JAEA-Technology-2014-038.pdf:3.84MB

The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.

JAEA Reports

Minutes of the IFMIF technical meetings; May 17-20, 2005, Tokyo, Japan

IFMIF International Team

JAERI-Review 2005-027, 416 Pages, 2005/08

JAERI-Review-2005-027.pdf:48.34MB

The International Fusion Materials Irradiation Facility (IFMIF) Technical Meetings were held on May 17-20, 2005 at Japan Atomic Energy Research Institute (JAERI) Tokyo. The main objectives were (1) to review technical status of the subsystems; accelerator, target and test facilities, (2) to technically discuss interface issues between target and test facilities, (3) to review results of peer-reviews performed in the EU and Japan, (4) to harmonize design / experimental activities among the subsystems, (5) to review and discuss the Engineering Validation and Engineering Design Activity (EVEDA) tasks, and (6) to make a report of (1) - (5) to the IFMIF Executive Subcommittee. This report presents a brief summary of the Target Technical Meeting, Test Facilities Technical Meeting, Target / Test Facilities Interface Meeting, Accelerator Technical Meeting and the Technical Integration Meeting.

Journal Articles

Analysis of benchmark results for reactor physics of LWR next generation fuels

Kitada, Takanori*; Okumura, Keisuke; Unesaki, Hironobu*; Saji, Etsuro*

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 8 Pages, 2004/04

Burnup calculation benchmark has been carried out for the LWR next generation fuels aiming at high burnup up to 70 GWd/t with UO$$_{2}$$ and MOX. Based on the submitted results by many benchmark participants, the present status of calculation accuracy has been confirmed for reactor physics parameters of the LWR next generation fuels, and the factors causing the calculation differences were analyzed in detail. Moreover, the future experiments and research subjects necessary to reduce the calculation differences were discussed and proposed.

Journal Articles

Validation of integrated burnup code system SWAT2 by the analyses of isotopic composition of spent nuclear fuel

Suyama, Kenya; Mochizuki, Hiroki*; Okuno, Hiroshi; Miyoshi, Yoshinori

Proceedings of International Conference on Physics of Fuel Cycles and Advanced Nuclear Systems; Global Developments (PHYSOR 2004) (CD-ROM), 10 Pages, 2004/04

This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models.

64 (Records 1-20 displayed on this page)