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

The Safety design guideline development for Generation-IV SFR systems

Nakai, Ryodai

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 10 Pages, 2017/06

The GIF Safety Design Criteria Task Force (SDC TF) has been developing a set of safety design guidelines (SDG) to support practical application of SDC since the completion of the "SDC Phase I Report" that clarifies safety design requirements for Gen-IV SFR systems. The main objective of the SDG development is to assist SFR developers and vendors to utilize the SDC in their design process for improving the safety in specific topical areas including the use of inherent/passive safety features and the design measures for prevention and mitigation of severe accidents. The first report on "Safety Approach SDGs" aims to provide guidance on safety approaches covering specific safety issues on fast reactor core reactivity and on loss of heat removal. The second report on "SDGs on key Structures, Systems and Components (SSCs)" focuses on the functional requirements for SSCs important to safety; reactor core system, reactor coolant system, and containment system.

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

no abstracts in English

Journal Articles

Study on In-Vessel Retention (IVR) of unprotected accident for fast reactor, 1; Overview of IVR evaluation in Anticipated Transient without Scram (ATWS)

Suzuki, Toru; Sogabe, Joji; Tobita, Yoshiharu; Sakai, Takaaki*; Nakai, Ryodai

Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06

Journal Articles

Evaluation of recriticality behavior in the material-relocation phase for Japan sodium-cooled fast reactor

Suzuki, Toru; Tobita, Yoshiharu; Nakai, Ryodai

Journal of Nuclear Science and Technology, 52(11), p.1448 - 1459, 2015/11

 Times Cited Count:6 Percentile:50.82(Nuclear Science & Technology)

Journal Articles

A Scenario of core disruptive accident for Japan sodium-cooled fast reactor to achieve in-vessel retention

Suzuki, Toru; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu; Nakai, Ryodai; Koyama, Kazuya*

Journal of Nuclear Science and Technology, 51(4), p.493 - 513, 2014/04

 Times Cited Count:55 Percentile:98.25(Nuclear Science & Technology)

As the most promising concept of SFRs, the JAEA has selected the advanced loop-type fast reactor, so-called JSFR. The safety design requirements of JSFR for design extension condition are the prevention of severe accidents and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, in particular, the In-Vessel Retention (IVR) against postulated Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the in-vessel cooling of core materials are evaluated so as to achieve IVR. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulations. This is an unprecedented approach to the construction of a CDA scenario, and is an effective method to objectively investigate the factors of IVR failure and design measures against them. It was concluded that mechanical/thermal failures of the reactor vessel could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.

Journal Articles

Safety design criteria for generation IV sodium-cooled fast reactor system

Nakai, Ryodai; Sofu, T.*

GIF Symposium Proceedings/2012 Annual Report of NEA, No.7141, p.35 - 43, 2013/00

Journal Articles

Evaluation of core disruptive accident for sodium-cooled fast reactors to achieve in-vessel retention

Suzuki, Toru; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu; Nakai, Ryodai; Koyama, Kazuya*

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 10 Pages, 2012/12

The JAEA has selected the advanced loop-type fast reactor JSFR as the most promising concept for the commercialization. The safety design requirements of JSFR for Design Extension Condition are the control of severe plant conditions, including the prevention of accident progression and the mitigation of severe-accident consequences. For the mitigation of severe-accident consequences, the In-Vessel Retention (IVR) against Core Disruptive Accidents (CDAs) is required. In order to investigate the sufficiency of these safety requirements, a CDA scenario should be constructed, in which the elimination of power excursion and the achievement of IVR are evaluated. In the present study, the factors leading to IVR failure were identified by creating phenomenological diagrams, and the effectiveness of design measures against them were evaluated based on experimental data and computer simulation. It was concluded that mechanical/thermal failures of the reactor vessel could be avoided by adequate design measures, and a clear vision for achieving IVR was obtained.

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

Development of safety design criteria for the Generation-IV Sodium-cooled Fast Reactor

Nakai, Ryodai; Okano, Yasushi; Kubo, Shigenobu

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 5 Pages, 2012/12

Journal Articles

Development of Level 2 PSA methodology for sodium-cooled fast reactors; Overview of evaluation technology development

Suzuki, Toru; Nakai, Ryodai; Kamiyama, Kenji; Seino, Hiroshi; Koyama, Kazuya*; Morita, Koji*

NEA/CSNI/R(2012)2, p.381 - 391, 2012/07

For the probabilistic safety assessment (PSA) of sodium-cooled fast reactors (SFRs), JAEA consolidated the analytical methodologies and technical basis for all phases/sequences to be evaluated in the Level 2 PSA. In addition to the existing computational codes such as SAS4A, SIMMER-III, DEBNET, ARGO and APPLOHS, JAEA newly developed MUTRAN and SIMMER-LT in order to evaluate the long term behaviors of the material-relocation in the degraded core. These tools enabled the systematic assessment for the in-vessel accident sequences. For the ex-vessel accident sequences, JAEA also improved CONTAIN/LMR taking into account the feature of SFRs and verified the analytical models utilizing the new experiments such as sodium-concrete reaction test. In addition, the technical basis for constructing event trees was compiled, in which the dominant factors having significant effects on the event progression were corresponded to the related experiments and analytical results.

Journal Articles

Design and assessment approach on advanced SFR safety with emphasis on the core disruptive accident issue

Nakai, Ryodai

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009), p.207 - 220, 2012/00

The concept of defense in depth (DiD) shall be applied to the safety design of advanced SFRs. Through the prevention, detection and control of accident, core disruptive accident (CDA) shall be excluded from Design Basis Events (DBEs). Considering that the SFR reactor core is not the most reactive configuration unlike in the LWRs, design measures to prevent and mitigate the consequences of CDA are being considered as provisions for beyond design basis events (BDBEs). To effectively meet the future nuclear energy system safety goals, advanced SFR designs should exploit passive safety features to increase safety margins and to enhance reliability. In particular the safety approach to eliminate the severe re-criticality will be highly desirable, because with this approach, severe accidents in SFRs can be simply regarded as similar to LWRs.

Journal Articles

Safety design approach of Japan sodium-cooled fast reactor

Nakai, Ryodai; Sakai, Takaaki; Okano, Yasushi

Proceedings of International Conference on Toward and Over the Fukushima Daiichi Accident (GLOBAL 2011) (CD-ROM), 7 Pages, 2011/12

This paper describes the safety design approach of Japan Sodium-cooled Fast Reactor (JSFR), which is designed in FaCT project. The basic safety approach is to satisfy the safety goal for next generation reactor taking into account characteristics of sodium fast reactor (SFR) with applying the concept of defence-in-depth (DiD). In addition to the provisions for DiD levels 1-3, design measures to prevent and mitigate core disruptive accident (CDA) are assigned as provisions for in DiD level 4. The re-criticality free/IVR concept makes less design impact on the containment and is essential for a large SFR commercialization.

Journal Articles

Safety principles and safety approaches for next generation sodium-cooled fast reactor

Okano, Yasushi; Sakai, Takaaki; Nakai, Ryodai

Proceedings of 2011 International Congress on Advances in Nuclear Power Plants (ICAPP '11) (CD-ROM), p.719 - 727, 2011/05

Journal Articles

Safety strategy of JSFR eliminating severe recriticality events and establishing in-vessel retention in the core disruptive accident

Sato, Ikken; Tobita, Yoshiharu; Konishi, Kensuke; Kamiyama, Kenji; Toyoka, Junichi; Nakai, Ryodai; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vassiliev, Y. S.*; et al.

Journal of Nuclear Science and Technology, 48(4), p.556 - 566, 2011/03

In the JSFR design, elimination of severe recriticality events in the Core Disruptive Accident (CDA) is intended as an effective measure to assure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the Initiating Phase selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, with introduction of Inner Duct on the other hand. The effectiveness of these measures are reviewed based on existing experimental data and evaluations performed with validated analysis tools. It is judged that the present JSFR design can exlude severe power burst events.

Journal Articles

Development of system based code, 1; Reliability target derivation of structures and components

Kurisaka, Kenichi; Nakai, Ryodai; Asayama, Tai; Takaya, Shigeru

Journal of Power and Energy Systems (Internet), 5(1), p.19 - 32, 2011/01

The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code by considering the safety point of view. In the new method, the reliability target is derived from the proposal to a quantitative safety goal that was published by the nuclear safety commission of Japan and the quantitative safety design requirements on the core damage frequency and the containment failure frequency that were determined in the Fast Reactor Cycle Technology Development Project by Japan Atomic Energy Agency, by utilizing analysis models of a probabilistic safety assessment (PSA). The present method was applied to determination of the reliability target of the structures and components which constitute the reactor cooling system in the Japanese sodium-cooled fast reactor. As a result, we confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determination of the reliability target associated with a random failure of the structures and components, and that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.

Journal Articles

Development of level 2 PSA methodology for sodium-cooled fast reactors, 1; Overview of evaluation technology development

Nakai, Ryodai; Suzuki, Toru; Kamiyama, Kenji; Seino, Hiroshi; Koyama, Kazuya*; Morita, Koji*

Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10

The evaluation technology of Level-2 PSA for Sodium-Cooled Fast Reactors (SFRs) was established in order to systematically assess the core damage sequences. In addition to the existing computational tools for Level-2 PSA, the computational tools, MUTRAN and SIMMER-LT were developed for core material relocation phase. Also the analytical models, CORCON and VANESA, were improved based on newly performed experiments for the ex-vessel phase taking into account the feature of SFRs. The technical information was compiled as technical database used in the construction and quantification of level-2 PSA event trees for SFRs. The technical basis was established for the Level-2 PSA for SFRs.

Journal Articles

Development of system based code, 1; Reliability target derivation of structures and components

Kurisaka, Kenichi; Nakai, Ryodai; Asayama, Tai; Takaya, Shigeru

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 10 Pages, 2010/05

The present paper describes a new method for determining the target value of structural reliability in the framework of the System Based Code by considering the safety point of view. In the new method, the reliability target is derived from the proposal to a quantitative safety goal that was published by the nuclear safety commission of Japan and the quantitative safety design requirements on the core damage frequency and the containment failure frequency that were determined in the Fast Reactor Cycle Technology Development Project by Japan Atomic Energy Agency, by utilizing analysis models of a probabilistic safety assessment (PSA). The present method was applied to determination of the reliability target of the structures and components which constitute the reactor cooling system in the Japanese sodium-cooled fast reactor. As a result, we confirmed that the present method combined with the PSA analysis model for internal initiating events is applicable to determination of the reliability target associated with a random failure of the structures and components, and that the method related to seismic initiating events can derive the target value of the occurrence frequency at which any of the important structures and components fails due to an earthquake.

Journal Articles

Recent status and trend of domestic and international fast reactor development

Nakai, Ryodai

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 50(5), p.318 - 323, 2008/05

In the 21st century, the importance of FBR development has been recognized taking into account the reduction of environment burden as well as the stable supply of resources. Recently foreign countries also recognize the usefulness of nuclear energy and become to proceed the fast reactor development actively. The international cooperation has been operated in the framework of international forum such as GIF and INPRO. The cooperation is being focused on the direction for the standardization of fast reactor concept. In Japan the research and development governance structure has been established and the FaCT project has been initiated aiming at the commercialization of fast reactor.

Journal Articles

Current status and future direction of INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles)

Omoto, Akira*; Moriwaki, Masanao*; Sugimoto, Jun; Nakai, Ryodai

Nihon Genshiryoku Gakkai-Shi, 49(2), p.89 - 111, 2007/02

no abstracts in English

Journal Articles

History and current status regarding the research and development of minor actinide recycling

Nagaoki, Yoshihiro; Nakai, Ryodai

Genshiryoku eye, 53(1), p.58 - 61, 2007/01

The technology which separates the minor actinide from the high level radioactive waste and utilizes it as the nuclear fuel called "Minor Actinide (MA) Recycling." It explained that the significance of MA recycling, the history and current status of R&D and the major challenges. From the next time, the manufacturing technology development of MA content nuclear fuel, the irradiation test of MA content nuclear test fuel in the experimental reactor JOYO and the future program to strive to commercialization will be shown.

64 (Records 1-20 displayed on this page)