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論文

Experimental study on debris bed characteristics for the sedimentation behavior of solid particles used as simulant debris

Shamsuzzaman, M.*; 堀江 達郎*; 浮池 亮太*; 神山 基紀*; 森岡 徹*; 松元 達也*; 守田 幸路*; 田上 浩孝; 鈴木 徹*; 飛田 吉春

Annals of Nuclear Energy, 111, p.474 - 486, 2018/01

 被引用回数:7 パーセンタイル:77.16(Nuclear Science & Technology)

Particle bed characteristics are experimentally investigated for the sedimentation and subsequent bed formation of solid particles, related to the coolability aspects in core-disruptive accidents. Presently a series of experiments with gravity driven discharge of solid particles into a quiescent water pool was performed to evaluate bed formation characteristic in the course of particle sedimentation. We evaluated the effects of the crucial factors: nozzle diameter, particle density, particle diameter and nozzle height on four key quantitative parameters of bed shape: mound dimple area, mound dimple volume, repose angle and mound height to illustrate the role of the crucial factors on forming the particle bed shape. The investigated crucial factors exhibit a significant role that diversifies the particle bed formation process. Based on the data obtained in the experimental observations, we developed an empirical correlation to compare the predicted results with the experimental bed heights. The proposed empirical correlation can reasonably demonstrate the general trend of the experimental bed height. This correlation could be useful to assess the particle bed elevation, and to identify the governing parameters.

論文

Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

小野田 雄一; 松場 賢一; 飛田 吉春; 鈴木 徹

Mechanical Engineering Journal (Internet), 4(3), p.16-00597_1 - 16-00597_14, 2017/06

For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident (ULOF) were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.

論文

高速炉の炉停止失敗事象における炉容器内終息(IVR)に関する検討,2; ULOF再配置/冷却過程における評価

曽我部 丞司; 鈴木 徹; 和田 雄作; 飛田 吉春

日本機械学会論文集(インターネット), 83(848), p.16-00393_1 - 16-00393_10, 2017/04

高速炉の代表的な炉停止失敗事象である冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)の熱的影響を評価するためには、事故が核的に終息した後の再配置/冷却過程において、損傷炉心物質が炉容器内のどこに再配置し、それぞれの場所で長期にわたって安定冷却できるかを示す必要がある。本報では、IVR(In-Vessel Retention)成立性に関する見通しを得るために実施した低圧プレナム移行燃料及び炉心残留燃料の安定冷却性評価について報告する。

論文

高速炉の炉停止失敗事象における炉容器内終息(IVR)に関する検討,1; ATWSにおけるIVR評価の概要

鈴木 徹; 曽我部 丞司; 飛田 吉春; 堺 公明*; 中井 良大

日本機械学会論文集(インターネット), 83(848), p.16-00395_1 - 16-00395_9, 2017/04

高速炉の炉停止失敗事象(ATWS: Anticipated Transient without Scram)に対して、原子炉容器内での事象終息(IVR: In-Vessel Retention)の成立性を検討した。検討においては、確率論的評価に基づいて冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)をATWSの代表事象に選定した上で、総合的安全解析コードや個別物理モデルを活用して炉心損傷時の事象進展を解析し、事故の機械的影響と熱的影響を評価した。本検討の結果から、原子炉容器は機械的にも熱的にも損傷することはなく、IVRが成立する見通しを得ることができた。

論文

Irradiation test of semiconductors components on the shelf for nuclear robots based on Fukushima Accidents

川妻 伸二; 中井 宏二; 鈴木 義晴; 加瀬 健

QST-M-2; QST Takasaki Annual Report 2015, P. 81, 2016/12

原子力施設の緊急時対応や廃止措置のためのロボットに使用される市販半導体素子の耐放射線性を評価した。福島第一原子力発電所事故の直後、市販半導体の耐放射線性評価と管理方法に関するガイドラインの作成が試みられた。その際に用いられたデータは、高放射線量かつ高汚染環境下で使用される力フィードバック型サーボマニピュレータ開発の一環として開発された古いデータベースであった。耐放射線性はかなり保守的に評価された。その理由は、主としてシリコンを母材とする古い半導体のデータであったためである。現在、ガリウム・ヒ素を簿在とする半導体が主流になりつつあり、耐放射線性もより高いと期待される。そのため、現在、市販されている半導体の照射試験を行い、耐放射線性の評価を行った。

論文

In-vessel retention of unprotected accident in a fast reactor; Assessment of material-relocation and heat-removal behavior in ULOF

曽我部 丞司; 鈴木 徹; 和田 雄作; 飛田 吉春

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

The achievement of in-vessel retention (IVR) of accident consequences in an unprotected loss of flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, is an effective and rational approach to enhancing the safety characteristics of the sodium-cooled fast reactor. The objective of the present study is to show that the decay heat generated from the relocated fuels would be stably removed in post-accident-material-relocation/post-accident-heat-removal (PAMR/PAHR) phase, where the relocated fuels mean fuel discharged from the core into low-pressure plenum through control-rod guide tubes, and fuel remnant in the disrupted core region (non-discharged fuel). As a result of the assessment, it can be concluded that the stable cooling of the relocated fuels was confirmed and the prospect of IVR was obtained.

論文

Improvements to the simmer code model for steel wall failure based on EAGLE-1 test results

豊岡 淳一; 神山 健司; 飛田 吉春; 鈴木 徹

Proceedings of 10th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-10) (USB Flash Drive), 7 Pages, 2016/11

In this paper, for the purpose establishing more generalized models for the SIMMER code to reproduce the effect of steel component on mixture-to-wall heat transfer in the EAGLE-1 program, the authors performed a model improvement for the SIMMER code to treat the direct contact of the molten steel in a more mechanistic manner. By this model improvement, evaluations with unifying agreement on a result of the EAGLE-1 program using the SIMMER code could be possible.

論文

An Empirical correlation to predict the distance for fragmentation of simulated Molten-Core materials discharged into a sodium pool

松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 8 Pages, 2016/10

ナトリウム冷却高速炉の炉心損傷時に原子炉容器下部プレナムへ流出した溶融炉心物質がデブリ化するまでの距離の評価を目的として、溶融炉心模擬物質を冷却材中へ放出させる試験を行い、デブリ化距離と流出条件の関係を実験相関式として整理した。実験相関式による予測は実験結果とよく一致した。本研究により、冷却材の沸騰・膨張によるデブリ化促進効果を考慮することで、ナトリウム中におけるデブリ化距離を適切に評価可能であることがわかった。

論文

高速炉の炉停止失敗事象における炉容器内終息(IVR)に関する検討,2; ULOF再配置/冷却過程における評価

曽我部 丞司; 鈴木 徹; 和田 雄作; 飛田 吉春

第21回動力・エネルギー技術シンポジウム講演論文集(USB Flash Drive), 3 Pages, 2016/06

高速炉の代表的な炉停止失敗事象(ATWS)である冷却材流量喪失時炉停止失敗事象(ULOF: Unprotected Loss of Flow)の再配置/冷却過程における事象推移を評価・検討し、IVR成立の見通しを得た。

論文

高速炉の炉停止失敗事象における炉容器内終息(IVR)に関する検討,1; ATWSにおけるIVR評価の概要

鈴木 徹; 曽我部 丞司; 飛田 吉春; 堺 公明*; 中井 良大

第21回動力・エネルギー技術シンポジウム講演論文集(USB Flash Drive), 4 Pages, 2016/06

The achievement of In-Vessel Retention (IVR) against Anticipated Transient without Scram (ATWS) is an effective and rational approach in enhancing the safety characteristics of sodium-cooled fast reactors. Based on the Probabilistic Risk Assessment (Level 1 PRA) for a prototype fast-breeder reactor, Unprotected Loss of Flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, can be selected as a representative event of ATWS. The objective of the present study is to show that no significant mechanical energy release would take place during core disruption caused by ULOF, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. As a result of the present evaluation with computational codes and physical models reflecting the knowledge on relevant experimental studies, the prospect of IVR against ULOF was obtained.

論文

Preliminary analysis of the post-disassembly expansion phase and structural response under unprotected loss of flow accident in prototype sodium cooled fast reactor

小野田 雄一; 松場 賢一; 飛田 吉春; 鈴木 徹

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 10 Pages, 2016/06

For the prototype sodium-cooled fast reactor, MONJU, the mechanical energy and structural response under energetics caused by neutronic power excursion during Unprotected Loss of Flow accident were preliminarily analyzed. The objective of this study is to demonstrate the integrity of the reactor vessel against the mechanical load induced by the energetics. Conservative energy production was assumed in order to confirm the robustness of the safety design of MONJU. Mechanical energy was evaluated with the code in which mechanistic modelling of core expansion was implemented. The mechanical energy, which were obtained by analyzing the expanding behavior of core materials after energetics, were about one order of magnitude below the thermodynamic work potential calculated by assuming isentropic expansion of the fuel vapor to one atmosphere, which was often used as an indicator to express the severity of the energetics. Structural integrity was then evaluated with coupled fluid-structure dynamics code using the obtained mechanical energy. No or very small circumferential residual strain of the reactor vessel was evaluated in most analytical cases, and even in the most conservative energy production case, the residual strain was only 0.008 % so that the integrity of the reactor vessel is maintained. The result obtained in the present study shows that MONJU has enough robustness against the mechanical load under energetics.

論文

Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

飛田 吉春; 神山 健司; 田上 浩孝; 松場 賢一; 鈴木 徹; 磯崎 三喜男; 山野 秀将; 守田 幸路*; Guo, L.*; Zhang, B.*

Journal of Nuclear Science and Technology, 53(5), p.698 - 706, 2016/05

AA2015-0794.pdf:2.46MB

 被引用回数:10 パーセンタイル:78.98(Nuclear Science & Technology)

炉心損傷事故(CDA)の炉内格納(IVR)はナトリウム冷却高速炉(SFR)の安全特性向上において極めて重要である。SFRのCDAにおいては、溶融炉心物質が炉容器の下部プレナムへ再配置し、構造物へ重大な熱的影響を及ぼし、炉容器の溶融貫通に至る可能性がある。この再配置過程の評価を可能とし、SFRのCDAではIVRで終息することが最も確からしいことを示すため、SFRのCDAにおける物質再配置挙動の評価手法を開発する研究計画が実施された。この計画では、炉心領域からの溶融物質流出挙動の解析手法、溶融炉心物質のナトリウムプール中への侵入挙動、デブリベッド挙動のシミュレーション手法を開発した。

論文

Lithium intercalation and structural changes at the LiCoO$$_{2}$$ surface under high voltage battery operation

田港 聡*; 平山 雅章*; 鈴木 耕太*; 田村 和久; 湊 丈俊*; 荒井 創*; 内本 喜晴*; 小久見 善八*; 菅野 了次*

Journal of Power Sources, 307, p.599 - 603, 2016/03

 被引用回数:26 パーセンタイル:73.4(Chemistry, Physical)

SrRuO$$_{3}$$(100)/Nb:SrTiO$$_{3}$$(100)上に作成したエピタキシャルLiCoO$$_{2}$$(104)薄膜を用いて、リチウムイオン電池正極材料の充放電過程における劣化機構について議論した。Li$$_{3}$$PO$$_{4}$$でLiCoO$$_{2}$$表面をコーティングしたものと、していないものを比較すると、コーティングすることで正極材料の劣化が抑えられることがわかった。表面X線回折実験で充放電過程における表面構造を追跡すると、コーティングした場合のLiCoO$$_{2}$$(104)表面は構造変化が少なく、その結果結晶構造が維持されて、劣化が抑えられていることが分かった。

報告書

平成26年度核燃料サイクル工学研究所放出管理業務報告書(排水)

渡辺 均; 中野 政尚; 藤田 博喜; 河野 恭彦; 井上 和美; 吉井 秀樹*; 大谷 和義*; 檜山 佳典*; 菊地 政昭*; 坂内 信行*; et al.

JAEA-Review 2015-030, 115 Pages, 2015/12

JAEA-Review-2015-030.pdf:25.28MB

本報告書は、原子力規制関係法令を受けた「再処理施設保安規定」、「核燃料物質使用施設保安規定」、「放射線障害予防規程」、「放射線保安規則」及び「茨城県等との原子力施設周辺の安全確保及び環境保全に関する協定書」、「水質汚濁防止法」並びに「茨城県条例」に基づき、平成26年4月1日から平成27年3月31日までの期間に日本原子力研究開発機構核燃料サイクル工学研究所から環境へ放出した放射性排水の放出管理結果をとりまとめたものである。再処理施設、プルトニウム燃料開発施設をはじめとする各施設からの放射性液体廃棄物は、濃度及び放出量ともに保安規定及び協定書等に定められた基準値を十分に下回った。

論文

A Numerical study on local fuel-coolant interactions in a simulated molten fuel pool using the SIMMER-III code

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Annals of Nuclear Energy, 85, p.740 - 752, 2015/11

 被引用回数:21 パーセンタイル:92.15(Nuclear Science & Technology)

Studies on local fuel-coolant interactions (FCI) in a molten pool are crucial to the analyses of severe accidents that could occur for sodium-cooled fast reactors (SFRs). To clarify the characteristics of this interaction, in recent years a series of simulated experiments, which covers a variety of conditions including much difference in water volume, melt temperature, water subcooling and water release site (pool surface or bottom), was conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, motivated by acquiring further evidence for understanding its mechanisms, interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency, are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is confirmed that, similar to experiments, the water volume, melt temperature and water release site are observable to have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the most probable reason leading to the limited pressurization and resultant mechanical energy release for a given melt and water temperature within the non-film boiling range, even under a condition of much larger volume of water entrapped within the pool, should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

論文

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

鈴木 徹; 飛田 吉春; 中井 良大

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

 被引用回数:5 パーセンタイル:48.17(Nuclear Science & Technology)

As the most promising concept of sodium-cooled fast reactors, the Japan Atomic Energy Agency (JAEA) has selected the advanced loop-type fast reactor, so-called JSFR. Through the evaluation of event progressions during hypothetical Core Disruptive Accident (CDA) under the design extension condition (DEC), a CDA scenario for JSFR has been evaluated. It has already been demonstrated that In-Vessel Retention (IVR) against CDA could be achieved by taking adequate design measures under best estimate conditions. The whole sequence of CDA scenario for JSFR was categorized into four phases according to the progress of core-disruption status. In the third phase, so-called material-relocation phase, the accident events would progress in the subcritical state. However, if the uncertainties about the molten state of core remnant and their discharge behavior outward from core are conservatively superposed, the disrupted core may lead up to recriticality. In the present study, the factors leading to recriticality in the material-relocation phase were investigated using the phenomenological diagrams, and the recriticality behaviors were evaluated through parametric analyses using SIMMER-III/IV codes. The results of parametric analyses suggested that a significant mechanical energy leading to the boundary failure of reactor vessel would not be released even assuming recriticality due to the uncertainties about molten state and discharge behavior. Through the present evaluation of the hypothetical recriticality event, the CDA scenario for JSFR could obtain further robustness from the viewpoint of achieving IVR.

論文

First analysis of local fuel-coolant interactions in a molten pool by SIMMER-III using reactor materials

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

To clarify the mechanisms underlying local fuel-coolant interactions (FCI) in a molten pool, in this study, several latest calculations with reactor materials were performed using SIMMER-III, an advanced fast reactor safety analysis code. The performed SIMMER-III analyses suggest that despite of a comparatively larger temperature range of molten-fuel and sodium possibly varied during reactor accidents, the isolation effect of vapor bubbles generated at the melt-sodium interface seems to be the unique dominant mechanism that leads to the limited pressurization. Knowledge and fundamental data from this work might be utilized for future empirical-approach studies (e.g. those investigating the characteristics of critical coolant volume required for achieving the saturated pressurization at varied melt and coolant temperatures).

論文

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

鈴木 徹; 飛田 吉春; 川田 賢一; 田上 浩孝; 曽我部 丞司; 松場 賢一; 伊藤 啓; 大島 宏之

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

 被引用回数:17 パーセンタイル:86.9(Nuclear Science & Technology)

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss-of-flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of In-Vessel Retention (IVR) for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of IVR against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.

論文

The Effect of coolant quantity on local fuel-coolant interactions in a molten pool

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Annals of Nuclear Energy, 75, p.20 - 25, 2015/01

 被引用回数:6 パーセンタイル:54.79(Nuclear Science & Technology)

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes.

論文

SIMMER-III analyses of local fuel-coolant interactions in a simulated molten fuel pool; Effect of coolant quantity

Cheng, S.; 松場 賢一; 磯崎 三喜男; 神山 健司; 鈴木 徹; 飛田 吉春

Science and Technology of Nuclear Installations, 2015, p.964327_1 - 964327_14, 2015/00

 被引用回数:2 パーセンタイル:22.25(Nuclear Science & Technology)

To clarify the mechanisms underlying local fuel-coolant interactions (FCI) in a molten pool, in recent years several experimental tests, with comparatively larger difference in coolant volumes, were conducted at the Japan Atomic Energy Agency by delivering a given quantity of water into a molten pool formed with a low-melting-point alloy. In this study, to further understand this interaction, interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are investigated using the SIMMER-III, an advanced fast reactor safety analysis code. It is found that the SIMMER-III code not only reasonably simulates the transient pressure and temperature variations during local FCIs, but also supports the limited tendency of pressurization and resultant mechanical energy release as observed from experiments when the volume of water delivered into the pool increases. The performed analyses also suggest that the most probable reason leading to such limited tendency should be primarily due to an isolation effect of vapor bubbles generated at the water-melt interface.

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