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Journal Articles

Validation study of initiating phase evaluation method for the core disruptive accident in an SFR

Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 10 Pages, 2019/05

Core Disruptive Accident (CDA) has been considered as one of the important safety issues in the severe accident evaluation of Sodium-cooled Fast Reactor (SFR), and SAS4A code is developed for Initiating Phase (IP) of CDA. Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to enhance its reliability in this study. SAS4A was validated in the following steps: (1) selection of the figure of merit (FOM) corresponding to Unprotected Loss Of Flow (ULOF) which is one of the most important and typical events in CDA, (2) identification of the phenomena involved in ULOF, (3) ranking the important phenomena, (4) development of the code validation test matrix, and (5) test analyses for validation corresponding to the test matrix. The reliability and validity of SAS4A code were significantly enhanced by this validation with PIRT approach.

Journal Articles

SAS4A simulations of selected CABRI-1 oxide fuel experiments

Karahan, A.*; Kawada, Kenichi; Tentner, A.*

Proceedings of 2018 ANS Winter Meeting and Nuclear Technology Expo; Embedded Topical International Topical Meeting on Advances in Thermal Hydraulics (ATH 2018) (USB Flash Drive), 4 Pages, 2018/11

Journal Articles

SAS4A analysis study on the initiating phase of ATWS events for generation-IV loop-type SFR

Kubota, Ryuzaburo; Koyama, Kazuya*; Moriwaki, Hiroyuki*; Yamada, Yumi*; Shimakawa, Yoshio*; Suzuki, Toru; Kawada, Kenichi; Kubo, Shigenobu; Yamano, Hidemasa

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04

This paper describes an analysis study on the initiating phase of the ATWS events with SAS4A in order to confirm the appropriateness of the core design for the medium-scale SFR (750MWe-1765MWt). Not using a conventional lumping method that multiple fuel sub-assemblies having a similar characteristic were assigned to one channel (representing fuel assembly in SAS4A), each channel represents only the sub-assemblies of identical operating condition. In addition, the detailed power and reactivity distribution were set reflecting the change of insertion position of control rods. Applying these detailed analysis conditions, the SAS4A analyses were performed for unprotected loss-of-flow (ULOF) and unprotected transient overpower (UTOP) during both of the nominal power and the partial power operation. As a result, more proper event progression including incoherency of events especially fuel dispersion after fuel failure was successfully evaluated and then this analysis study suggested that the power excursion with prompt criticality leading to large mechanical energy release can be prevented in the initiating phase of the current design.

Journal Articles

Validation study in SAS4A code in simulated mild TOP condition

Kawada, Kenichi; Suzuki, Toru

Transactions of the American Nuclear Society, 115(1), p.1597 - 1598, 2016/11

Journal Articles

A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 Times Cited Count:11 Percentile:10.49(Nuclear Science & Technology)

Journal Articles

Preliminary result of validation study in SAS-SFR (SAS4A) code in simulated top and undercooled overpower conditions

Kawada, Kenichi; Takahashi, Katsuhiko*; Tobita, Yoshiharu

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 10 Pages, 2014/12

Journal Articles

Development of PIRT (phenomena identification and ranking table) for SAS-SFR (SAS4A) validation

Kawada, Kenichi; Sato, Ikken; Tobita, Yoshiharu; Pfrang, W.*; Buffe, L.*; Dufour, E.*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07

Journal Articles

Safety evaluation of prototype fast-breeder reactor; Analysis of ULOF accident to demonstrate in-vessel retention

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Ito, Kei

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 10 Pages, 2014/07

Journal Articles

CAF$'E$ experiments on the flow and freezing of metal fuel and cladding melts, 1; Test conditions and overview of the results

Fukano, Yoshitaka; Kawada, Kenichi; Sato, Ikken; Wright, A. E.*; Kilsdonk, D. J.*; Aeschlimann, R. W.*; Bauer, T. H.*

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 10 Pages, 2012/00

Journal Articles

CAF$'E$ experiments on the flow and freezing of metal fuel and cladding melts, 2; Results, analysis, and applications

Wright, A. E.*; Bauer, T. H.*; Kilsdonk, D. J.*; Aeschlimann, R. W.*; Fukano, Yoshitaka; Kawada, Kenichi; Sato, Ikken

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles (FR 2009) (CD-ROM), 9 Pages, 2012/00

Journal Articles

Halo structure of the island of inversion nucleus $$^{31}$$Ne

Nakamura, Takashi*; Kobayashi, Nobuyuki*; Kondo, Yosuke*; Sato, Yoshiteru*; Aoi, Nori*; Baba, Hidetada*; Deguchi, Shigeki*; Fukuda, Naoki*; Gibelin, J.*; Inabe, Naoto*; et al.

Physical Review Letters, 103(26), p.262501_1 - 262501_4, 2009/12

 Times Cited Count:142 Percentile:2.51(Physics, Multidisciplinary)

no abstracts in English

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.

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 degree 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

Safety Characteristics of Mid-sized MOX Fueled Liquid Metal Reactor Core of High Converter Type in the Initiating Phase of Unprotected Loss of Flow Accident; Effects of low specific fuel power density on ULOF behavior brought by employment of large diameter fuel pins

Ishida, Masayoshi; Kawada, Kenichi; Niwa, Hajime

JNC-TN9400 2003-059, 74 Pages, 2003/07

JNC-TN9400-2003-059.pdf:1.58MB

Safety characteristics in core disruptive accidents (CDAs) of mid-sized MOX fueled liquid metal reactor core of high converter type have been examined by using the CDA initiating phase analysis code SAS4A. The design concept of high converter type reactor core has been studied as one of options in the category of sodium-cooled reactor in Phase II of Feasibility Study on Commercialized Fast Reactor Cycle System.An unprotected loss-of-flow accident (ULOF) has been selected as a representative CDA initiator for this study. A core concept of high converter type, which employed a large diameter fuel pin of 11.1mm with 1.2m core height to get a large fuel volume fraction in the core to achieve high internal conversion ratio was proposed in JFY2001. Each fuel subassembly of the core (abbreviated here as UPL120) was provided with an upper sodium plenum directly above the core to reduce the sodium void reactivity worth. Because of the large fuel pin diameter, average specific fuel power density (31 kW/kg-MOX) of UPL120 is about one half of those of conventional large MOX cores. The reactivity worth of sodium voiding is 6$ in the whole core, and -1$ in the all upper plenums. Initiating phase of ULOF accident in UPL120 under the conditions of nominal design and best estimate analysis resulted in a slightly super-prompt critical power burst. The causes of the super-prompt criticality have been identified twofold: (a) the low specific fuel power density of core reduced the effectiveness of prompt negative reactivity feedback of Doppler and axial fuel expansion effects upon increase in reactor power, and (b) the longer core height compared with conventional 1m cores brought, together with the lower specific power density, a remarkable delay in insertion of negative fuel dispersion reactivity after the onset of fuel disruption in sodium voided subassembly due to the lower linear heat rating in the top portion of the core. During the delay, burst-type fuel failures in sodium un-v

JAEA Reports

An evaluation study on ULOF event sequences in the prototype FBR; An evaluation of CDA reflecting the latest knowledge

Tobita, Yoshiharu; Morita, Koji; Kawada, Kenichi; Niwa, Hajime; Nonaka, Nobuyuki

PNC-TN9410 97-079, 106 Pages, 1997/09

PNC-TN9410-97-079.pdf:3.9MB

The sequences of ULOF (unprotected loss-of-flow) event in the prototype FBR has been evaluated, as a part of the research and development (R&D) in the reactor safety research, reflecting the latest experimental and analytical knowledge on CDA (core disruptive accident) which has been accumulated at O-arai Engineering Center. In the R&D activity on the FBR reactor safety subject, we have accumulated the experimental knowledge of mitigation mechanism in the energy generating process in CDA, utilizing international in-pile safety experimental programs such as CABRI program, as well as the out-of-pile experiments in Japan and foreign countries. This knowledge has been reflected to the development and validation of the SAS and SIMMER code. The objectives of this study are to apply these new assessment techniques to the prototype FBR and to clarify quantitatively in detail the energy generation process of CDA. In this study, an emphasis is placed on the event sequence of the melt progression phase ("transition phase") which has been recognized as one of the important issues of CDA analysis. The major parameters to be considered in this phase are the change of the mobile molten fuel mass and the history of the fuel motion, and also the relation between these parameters and energy generation mechanism. The following methods and approaches have been taken into account in this evaluation study. (a)The SAS4A code is used for the analysis of the transient behavior in the first Phase driven by core voiding ("initiating phase"), and the SIMMER-III code is used for the latter phases with melt-progression (tansition phase) and also the energy conversion process from the thermal one to the mechanical one. These codes have been developed and validated under the collaboration among PNC, CEA and FZK. (b)The uncertainty band of the void reactivity worth and Doppler coeficient has been reduced through the re-evaluation of the critical experimental data in the neutron physics area. ...

Oral presentation

Probabilistic safety assessment on experimental fast reactor Joyo, 3-2; Evaluation of UTOP initiating phase for Joyo

Kawada, Kenichi; Fukano, Yoshitaka; Sato, Ikken

no journal, , 

no abstracts in English

Oral presentation

Probabilistic safety assessment on the experimental fast reactor JOYO, 4-2; Assessment of the event progression of core disruption during UTOP event in JOYO

Tobita, Yoshiharu; Sato, Ikken; Kawada, Kenichi; Fukano, Yoshitaka

no journal, , 

The event progression of core disruption in UTOP (Unprotected Transient Overpower), which was judged to be an important core disruption category in risk assessment of JOYO, was analyzed. It was confirmed that power burst did not occur in the initial phase of the accident and was not likely to occur in the successive phase of disruption extension.

Oral presentation

Present status of the safety evaluation for the CDA mitigation in fast reactor, 1; Analysis of initiating phase in ULOF

Kawada, Kenichi; Ishida, Shinya; Onoda, Yuichi; Tobita, Yoshiharu

no journal, , 

no abstracts in English

Oral presentation

Interpretation of the CABRI-2 E12 test and SAS4A evaluation of the postfailure fuel behavior

Ishida, Shinya; Onoda, Yuichi; Kawada, Kenichi

no journal, , 

no abstracts in English

Oral presentation

24 (Records 1-20 displayed on this page)