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Aoyagi, Mitsuhiro; Sonehara, Masateru; Ishida, Shinya; Uchibori, Akihiro; Kawada, Kenichi; Okano, Yasushi; Takata, Takashi
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09
Kawada, Kenichi; Suzuki, Toru*
Nuclear Engineering and Technology, 53(12), p.3930 - 3943, 2021/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, a simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.
Sonehara, Masateru; Uchibori, Akihiro; Aoyagi, Mitsuhiro; Kawada, Kenichi; Takata, Takashi; Ohshima, Hiroyuki
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2021/07
In sodium-cooled fast reactors (SFRs), it has been pointed out that molten fuel may be discharged from the core during a severe accident (SA) accompanied by core damage, and may solidify into debri particles with diameters ranging from several millimeters to several hundred micrometers due to interaction with the sodium coolant and accumulate at the bottom of the reactor vessel. Therefore, it is necessary to understand the behavior of such debri particles appropriately to evaluate the SA event progression. To meet these requirements, a molten fuel behavior analysis code using dissipative particle dynamics (DPD), a kind of particle method, has been developed as a part of the SPECTRA code, tool for consistent analysis of in-vessel and ex-vessel events in sodium fast reactor accidents. In this study, it was found that the new analyses code can reproduce sedimentation behavior of particles by adding a new stress term in the shear direction.
Kawada, Kenichi; Suzuki, Toru*
Journal of Nuclear Science and Technology, 58(3), p.347 - 360, 2021/03
Times Cited Count：1 Percentile：12.16(Nuclear Science & Technology)SAS4A is presently the worldwide standard computer code for simulation of the initiating phase of the Core Disruptive Accident (CDA) in MOX-fueled Sodium-cooled Fast Reactors (SFR). In order to improve the capability of the computational code, SAS4A, to be applied to the safety assessment for SFRs, detailed investigations of physical phenomena under unprotected loss-of-flow (ULOF) conditions were conducted, and physical models to be developed and improved were identified using the Phenomena Identification and Ranking Table (PIRT) method. A fuel stub motion, in which the residual fuel pellets would move to the core central region after fuel pin disruption, was selected as one of the key phenomena to be simulated properly and a key concept of the fuel stub motion model was presented.
Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Mechanical Engineering Journal (Internet), 7(3), p.19-00523_1 - 19-00523_17, 2020/06
The Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to indicate the reliability of SAS4A code sufficiently and objectively. Based on this approach, issue and objective were clarified, plant design and scenario were defined, FOM and key phenomena were selected, and the code validation test matrix was completed with the results of investigation about analysis models and test cases. The results of the test analysis corresponding to this matrix show that the SAS4A models required for the IP evaluation were sufficiently validated. Furthermore, the validation with this matrix is highly reliable, since this matrix represents the comprehensive validation that also considers the relation between physical phenomena. In this study, the reliability and validity of SAS4A code were significantly enhanced by using PIRT approach to the sufficient level for CDA analyses in 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.
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
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.
Kawada, Kenichi; Suzuki, Toru
Transactions of the American Nuclear Society, 115(1), p.1597 - 1598, 2016/11
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：27 Percentile：91.4(Nuclear Science & Technology)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
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
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
experiments on the flow and freezing of metal fuel and cladding melts, 1; Test conditions and overview of the resultsFukano, 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
experiments on the flow and freezing of metal fuel and cladding melts, 2; Results, analysis, and applicationsWright, 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
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.
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
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.
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
; Tobita, Yoshiharu; ; ; Ishida, Masayoshi; ;
JNC TN9400 2001-056, 64 Pages, 2001/03
JNC-TN9400-2001-056.pdf:2.66MB
The Phase I of the Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System is being performed for two years from Fisca1 Year (FY) 1999. This report describes the results obtained in FY 1999 as an interim report of the Phase I from the viewpoint of reactor safety for various FBR plant condidates. The objectives of the study are to understand the safety charaeteristics of advanced fuel and to examine the fulfillment of the target level of reactor safety in each plant concept. The items studied are the recriticality characteristics of degraded core for various core concepts, investigation of the measures for avoiding recriticality event, safety analysis of sodium cooled MOX fueled cores, target of void worth in core design for sodium cooled reactors, and investigation of core disruptive accident sequences in various reactor concepts. The results of this study have been reflected properly to the core and plant design. In FY 2000, the study will be continued along with the progress of the plant design in order to prepare for the judgment of the candidates from the viewpoint of reactor safety.
