Refine your search:     
Report No.
 - 
Search Results: Records 1-20 displayed on this page of 25

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

Research & development of safety approach and safety assessment for the next generation SFR

Okano, Yasushi; Kurisaka, Kenichi; Yamano, Hidemasa; Fujita, Satoshi; Nishino, Hiroyuki; Sakai, Takaaki

Proceedings of Probabilistic Safety Assessment and Management Topical Conference; In light of the Fukushima Dai-ichi Accident (PSAM 2013) (USB Flash Drive), 6 Pages, 2013/04

Journal Articles

Improved safety approach for general safety designs of the next generation sodium-cooled fast reactor systems

Okano, Yasushi; Yamano, Hidemasa; Fujita, Satoshi; Kubo, Shigenobu; Sakai, Takaaki; Nakai, Ryodai

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 9 Pages, 2012/12

General safety approaches are developed for next generation SFR based on the fundamental safety characteristics with incorporating lessons learned from the TEPCO's Fukushima Daiichi accidents. The fundamental characteristics are: reactivity, coolant pressure, sub-cool margin, ultimate heat sink, and sodium properties. These points are considered to derive general safety approach related to fundamental function. The key is to apply passive safety for prevention/mitigation of severe accident in design extension condition (DEC) with balancing active safety systems - passive mechanism should be built-in design for reactor shutdown and decay heat removal especially for DEC in order to enhance diversity to the engineered safety systems utilized for design basis accident. For containment integrity, the potentials of pressure/temperature increases via sodium leak and of significant mechanical energy release by re-criticality in the course of the CDA should be eliminated.

Journal Articles

A Three-dimensional neutronics-thermohydraulics simulation of core disruptive accident in sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi

Nuclear Engineering and Design, 239(9), p.1673 - 1681, 2009/09

 Times Cited Count:14 Percentile:69.01(Nuclear Science & Technology)

The SIMMER-III code is a two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. Since SIMMER-III is expected to become a standard tool for fast reactor safety analysis with likely application to licensing calculations, the code must be demonstrated to be sufficiently robust and reliable. For this purpose, a systematic assessment program of the code has been conducted in cooperation with European partners. The development of SIMMER-IV has been performed to be a direct extension of SIMMER-III to three dimensions. Recently, the parallelization of SIMMER-IV has been made to allow applications to reactor calculation within available computational resource. A three-dimensional calculation with SIMMER-IV is presented to indicate more realistic accident scenario.

Journal Articles

First 3-D calculation of core disruptive accident in a large-scale sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Maschek, W.*

Annals of Nuclear Energy, 36(3), p.337 - 343, 2009/04

 Times Cited Count:16 Percentile:71.14(Nuclear Science & Technology)

The SIMMER-IV computer code is a three-dimensional fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron transport kinetics model. The present study attempted the first application of SIMMER-IV to a core disruptive accident in a large-scale sodium-cooled fast reactor. A main point of this study was to investigate reactivity effects with fuel relocation under three-dimensional core representation including control rods. The calculation has indicated that the fuel discharge from the core was disturbed by a significant flow resistance at the entrance nozzle in the current design. Additional static neutronic calculations have been performed to compare basic neutronic characteristics between different scale cores. The static neutronic calculations have clarified that the outward fuel compaction within the inner core increased the reactivity in the large-scale core unlike the small-scale core.

Journal Articles

First 3-D calculation of core disruptive accident in a large scale sodium-cooled fast reactor

Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Maschek, W.*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 8 Pages, 2008/09

The present study attempted the first application of SIMMER-IV to a CDA in a large-scale SFR to draw event progression and to grasp key characteristics. Since the 3-D calculation requires much computation time, the SIMMER-IV calculation focused on the early stage of the transition phase in this study. The calculation result indicated mild event progression without recriticality. Compared to a small-scale SFR, it was found that the radial sloshing reactivity was not so significant in a large-scale SFR.

Journal Articles

Development of a three-dimensional CDA analysis code; SIMMER-IV and its first application to reactor case

Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Sato, Ikken; Niwa, Hajime

Nuclear Engineering and Design, 238(1), p.66 - 73, 2008/01

 Times Cited Count:31 Percentile:87.02(Nuclear Science & Technology)

For the transition phase analysis of core disruptive accidents, the development of a three-dimensional reactor safety analysis code, SIMMER-IV, has been carried out based on the technology of the two-dimensional SIMMER-III code. The world first application of SIMMER-IV to a small-sized sodium-cooled fast reactor has also been attempted to clarify event progression in the early stage of the transition phase. This SIMMER-IV calculation is compared to the two-dimensional case calculated by SIMMER-III, neglecting the presence of control rod guide tubes. The present analysis with the three-dimensional representation suggests that the conventional scenario leading to rather early high-mobility fuel-pool formation is unrealistic and the degraded core tends to keep low mobility in the early stage of transition phase.

JAEA Reports

Analysis of ULOF accident in Monju reflecting the knowledge from CABRI in-pile experiments and others

Sato, Ikken; Tobita, Yoshiharu; Suzuki, Toru; Kawada, Kenichi; Fukano, Yoshitaka; Fujita, Satoshi; Kamiyama, Kenji; Nonaka, Nobuyuki; Ishikawa, Makoto; Usami, Shin

JAEA-Research 2007-055, 84 Pages, 2007/05

JAEA-Research-2007-055.pdf:16.66MB

In the first licensing procedure of the prototype FBR "Monju", the event sequence of ULOF (Unprotected Loss of Flow) was analyzed and the estimated mechanical energy was about 380 MJ as an isentropic expansion potential to atmospheric pressure. The prototype FBR has been stopped for more than 10 years since the sodium leakage accident in the secondary loop in 1995. The neutronic characteristics of reactor core changed as a consequence of radioactive decay of fissile Plutonium during this shutdown period. In order to assess the effect of this neutronic characteristics change to the mechanical energy release in ULOF, the event sequence of ULOF was analyzed reflecting the current knowledge, which was obtained by safety studies after the first licensing of the prototype reactor. It was shown that the evaluated mechanical energy release became smaller than 380 MJ, even with the change of neutronic characteristics.

Journal Articles

SIMMER-III; A Coupled neutronics-thermohydraulics computer code for safety analysis

Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Suzuki, Toru; Kamiyama, Kenji; Morita, Koji*; Maschek, W.*; Pigny, S.*

Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 8 Pages, 2007/04

To simulate complex phenomena during core disruptive accidents in sodium-cooled fast reactors, JAEA has been developing the SIMMER-III code,which is two-dimensional, multi-velocity-field, multi-phase, multi-component, Eulerian, fluid-dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. Recently, the three-dimensional code SIMMER-IV is also developed with the same physical model as SIMMER-III. In the present paper, the models and methods of SIMMER-III/IV are briefly reviewed with highlighting the recent improvements. The major achievements of the code assessment program are then described, followed by presentation of practical applications. A three-dimensional calculation with SIMMER-IV are also shown to indicate more realistic accident scenario. In addition, this calculation result show the disrupted core state for investigating the post-accident material relocation and heat removal phase.

Journal Articles

Development of a three-dimensional CDA analysis code; SIMMER-IV, and its first application to reactor case

Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Sato, Ikken; Niwa, Hajime

Proceedings of Technical Meeting on Severe Accident and Accident Management (CD-ROM), 12 Pages, 2006/03

For the transition phase analysis of core disruptive accidents, the development of a three-dimensional reactor safety analysis code, SIMMER-IV, has been carried out based on the technology of the two-dimensional SIMMER-III code. The world first application of SIMMER-IV to a small-sized sodium-cooled fast reactor was also attempted to clarify event progression in the early stage of the transition phase. This SIMMER-IV calculation was compared to the two-dimensional case calculated by SIMMER-III, neglecting the presence of control rod guide tubes. The present analysis with the three-dimensional representation suggested that the conventional scenario leading to rather early high-mobility fuel-pool formation is unrealistic and the degraded core tends to keep low mobility in the early stage of transition phase.

JAEA Reports

Study on countermeasures for the elimination of re-criticality issue for the sodium cooled reactors; Results of the Studies in 2003

Kubo, Shigenobu; Tobita, Yoshiharu; Kawada, Kenichi; Onoda, Yuichi; Sato, Ikkenn; Kamiyama, Kenji; Ueda, Nobuyuki*; Fujita, Satoshi; Niwa, Hajime

JNC TN9400 2004-041, 135 Pages, 2004/07

JNC-TN9400-2004-041.pdf:17.3MB

This report shows the results of the study on countermeasures for the elimination of re-criticality issue for the sodium cooled reactors, which was conducted in 2003 as a part of the feasibility study phase II for the commercialization of fast reactors. A sort of analytical studies related to the in-vessel retention capability under the unprotected loss of flow condition was conducted for the large scale and medium scale sodium cooled reactors, aiming at establishing some promising concepts to resolve the re-criticality issue keeping consistency with the basic concept of the core and plant design. Major conclusions are as follows. ABLE concept, which is proposed as a measure to enhance the fuel discharge capability in the early transition phase, needs much time to initiate fuel discharge than wrapper tube failure. Therefore it is currently concluded that it is difficult to show clear perspective. A modified version of FAIDUS which has less drawbacks on the core and cycle performance and related R&Ds than original FAIDUS was proposed for further study. In-place retention and cooling in the core region is important from view point of reduction of R&D loads conceming post accident material relocation and cooling at the bottom of the reactor vessel. A possibility of which the in-vessel retention can be achieved by quantitatively clarifying the effect of the superior cooling potential of sodium was shown. Based on the currently available information related to FAIDUS and ABLE, possible candidates of experimental studies were shown. An initiating phase analysis for the metallic fuel core with 550$$^{circ}$$C of core outlet temperature and 8${$}$ of sodium void worth resulted in mild consequence without prompt criticality. Although there is still large uncertainty in the early transition phase, it might be possible to avoid severe re-criticality. And it was shown that power excursion due to molten fuel sloshing might be milder than that of MOX fuel case.

JAEA Reports

SIMMER-III: A Computer Program for LMFR Core Disruptive Accident Analysis; Version 3.A Model Summary and Program Description

Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kamiyama, Kenji; Kondo, Satoru; Morita, Koji*; Fischer, E. A.; Brear, D. J.; Shirakawa, Noriyuki*; Cao, X.; et al.

JNC TN9400 2003-071, 340 Pages, 2003/08

JNC-TN9400-2003-071.pdf:1.54MB

An advanced safety analysis computer code, SIMMER-III, has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs). SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a space-dependent neutron kinetics model. By completing and integrating all the physical models originally intended at the beginning of this code development project, SIMMER-III is now applicable to integral reactor calculations and other complex multiphase flow problems. A systematic code assessment program, conducted in collaboration with European research organizations, has shown that the advanced features of the code have resolved many of the limitations and problem areas in the previous SIMMER-II code. In this report, the models, numerical algorithms and code features of SIMMER-III Version 3.A are described along with detailed program description. Areas which require future model refinement are also discussed. SIMMER-III Version 3.A, a coupled fluid-dynamics and neutronics code system, is expected to significantly improve the flexibility and reliability of LMFR safety analyses.

JAEA Reports

SIMMER-IV: A Three-Dimensional Computer Program for LMFR Core Disruptive Accident Analysis; Version 2.A Model Summary and Program Description

Yamano, Hidemasa; Fujita, Satoshi; Tobita, Yoshiharu; Kondo, Satoru; Morita, Koji*; Sugaya, Masaaki*; Mizuno, Masahiro*; Hosono, Seigo*; Kondo, Teppei*

JNC TN9400 2003-070, 333 Pages, 2003/08

JNC-TN9400-2003-070.pdf:1.35MB

An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 2) coupled with the three-dimensional neutronics model. With the completion of the SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV. This report describes SIMMER-IV Version 2.A, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users.

JAEA Reports

CDA Analysis of lead-cooled fast reactor; Results in 2000

Tobita, Yoshiharu; ; Fujita, Satoshi

JNC TN9400 2001-050, 4 Pages, 2001/03

JNC-TN9400-2001-050.pdf:1.68MB

The feasibility study for the commercialization of fast reactors is underway in Japan Nuclear Cycle Development Institute, aiming at the achievement of the economic competitiveness, making full use of the natural resources, reduction of the environmental impact and the assurance of the nuclear non-proliferation and safety. This report shows the results of the analysis of the core-disruptive accident in lead-cooled reactor, and therby discusses the safety characteristics of the heavy metal cooled fast reactors. The analysis showed that the reactivity increase due to the molten clad relocation and fission gas blowdown was mild and did not lead to the recriticality. On the other hand, it was shown that the motion of disrupted fuel particles in the single phase lead-coolant had a possibility to produce recriticality. In addition, the importance of the integrity of the primary boundary, structures in the reactor vessel, and decay heat removal system against the high temperature lead was pointed out from the viewpoint of the in-vessel retention of the accident consequences.

Journal Articles

Mechanistic SIMMER-3 Analyses of Severe(ADS)Transients in Accelerator Driven Systems

Yamano, Hidemasa; ; ; Fujita, Satoshi

Proceedings of 9th International Conference on Nuclear Engineering (ICONE-9) (CD-ROM), 0 Pages, 2001/00

None

JAEA Reports

SIMMER-IV: A Three-Dimensional Computer Program for LMFR Core Disruptive Accident Analysis - Version 1.B Model Summary and Program Description -

kondo, Satoru; Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Morita, Koji*; Mizuno, Masahiro*; *

JNC TN9400 2001-003, 307 Pages, 2000/11

JNC-TN9400-2001-003.pdf:8.33MB

An advanced safety analysis computer code, SIMMER-III, has been developed at Japan Nuclear Cycle Development Institute (JNC) to more realistically investigate postulated core disruptive accidents in liquid-metal fast reactors. The two-dimensional framework of SIMMER-III fluid dynamics has been extended to three dimensions to a new code, SIMMER-IV, which is currently (in Version 1) coupled with the existing two-dimensional neutronics model. With the completion of the first SIMMER-IV version, the applicability of the code is further enhanced and the many of the known limitations in SIMMER-III are eliminated. The sample calculations demonstrated the general validity of SIMMER-IV. This report describes SIMMER-IV version 1.B, by documenting the models, numerical algorithms and code features, along with the program description and input and output information to aid the users. Further extension of the code is planned to couple the three-dimensional neutronics in the future.

JAEA Reports

SIMMER-III: A Computer Program for LMFR Core Disruptive Accident Analysis - Version 2. H Model Summary and Program Description -

kondo, Satoru; Yamano, Hidemasa; Suzuki, Toru; Tobita, Yoshiharu; Fujita, Satoshi; ; Kamiyama, Kenji

JNC TN9400 2001-002, 318 Pages, 2000/11

JNC-TN9400-2001-002.pdf:8.66MB

An advanced safety analysis computer eode, SIMMER-III, has been developed to investigate postulated core disruptive accidents in liquid-metal fast reactors (LMFRs). SIMMER-III is a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid--dynamics code coupled with a space-dependent neutron kinetics model. By completing and integrating all the physical models originally intended at the beginning of this code development project, SIMMER-III is now applicable to integral reactor calculations and other complex multiphase flow problems. A systematic code assessment program, conducted in collaboration with European research organizations, has shown that the advanced features of the code have resolved many of the limitations and problem areas in the previous SIMMER-II code. In this report, the models, numerical algorithms and code features of SIMMER-III version 2.H are described along with detailed program description. Areas which require future model refinement are also discussed. SIMMER-III Version 2.H, a coupled fluid-dynamics and neutronics code system, is expected to significantly improve the flexibility and reliablity of LMFR safety analyses.

JAEA Reports

Phase 2 code assessment of SIMMER-III; A computer program for LMFR core disruptive accident analysis

kondo, Satoru; Yamano, Hidemasa; Tobita, Yoshiharu; Fujita, Satoshi; Kamiyama, Kenji; W.Maschek*; P.Coste*

JNC TN9400 2000-105, 777 Pages, 2000/09

JNC-TN9400-2000-105.pdf:33.07MB

The liquid-metal fast reactor (LMFR) safety analysis computer code SIMMER-III successfully reached its first milestone with development of Version 2, a two-dimensional, three-velocity-field, multiphase, multicomponent, Eulerian, fluid-dynamics code coupled with a spase-dependent neutron kinetics model. The development and assessment of SIMMER-III has been participated internationally by Forschungszentrum Karlsruhe, Germany and Commissariat $'a$ l'Energie Atomique, France. To advance the code as a next-generation standard tool for LMFR safety analysis, it was agreed among the partners that a joint code assessment program should be conducted comprehensively and systematically. This program consists of a two-step effort: Phase 1 for fundamental or separate-effect assessment of individual code models; and Phase 2 for integral assessment of key physical phenomena relevant to LMFR safety. THis report describes the achivement of the SIMMER-III Phase 2 assessment program. The report details the results of each of the 34 test problems analyzed, conducted in five major categories of key LMFR safety phenomena, and synthesizes the outcome of the analyses. Through this extensive study, effectively utilizing the international collaboration scheme and world experimental database available, the SIMMER-III code has proved to be basically valid both numerically and physically, with significantly enhanced applicability and flexibility over its predecessor, SIMMER-II. Thus the code is now applicable to integral reactor safety calculations. The study has also identified limitations and problem areas on which future code development and assessment should focus.

JAEA Reports

CDA analysis of lead-cooled fast reactor

Tobita, Yoshiharu; ; Fujita, Satoshi

JNC TN9400 2000-082, 24 Pages, 2000/07

JNC-TN9400-2000-082.pdf:0.92MB

The feasibility study for the commercialization of fast reactors is underway in Japan Nuclear Cycle Development Institute, aiming at the achievement of the economic competitiveness, making full use of the natural resources, reduction of the environmental impact and the assurance of the nuclear non-proliferation and safety. This report shows the results of the analysis of the core-disruptive accident in lead-cooled reactor, which was performed to clarify the safety aspects of the heavy metal cooled fast reactors. The analysis showed that the high boiling point and density of lead made the event progression in the core-disruptive accident genuine and mild and prohibited the energetic re-criticality. Therefore, the dedicated designs to avoid the energetic recriticality, such as the fuel sub-assembly with inner duct and/or the fuel sub-assembly with partial removal of axial blanket pellet are not necessary in the lead-cooled fast reactor. On the other hand, the importance of the integrity of the primary boundary, structures in the reactor vessel, and decay heat removal system against the high temperature lead was pointed out from the viewpoint of the in-vessel retention of the accident.

Journal Articles

SPACE-TIME KINETICS SIMULATION OF AN EARLY BURST PHASE OF THE CRITICALITY ACCIDENT

; Fujita, Satoshi; ; Yamano, Hidemasa

International Workshop on the Safety of the Nuclear Fuel Cycle, 0 Pages, 2000/00

None

Journal Articles

CURRENT STATUS AND APPLICATION OF SIMMER-3,AN ADVANCED COMPUTER PROGRAM FOR LMFR SAFETY ANALYSIS

; Fujita, Satoshi; ; Yamano, Hidemasa

Pro.65-72, p.65 - 72, 2000/00

None

25 (Records 1-20 displayed on this page)