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

http://doi.org/10.1299/mej.16-00597

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.

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

https://doi.org/10.1299/transjsme.16-00395

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。

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.

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。

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.

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

https://doi.org/10.1080/00223131.2016.1143409

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。

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。

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

https://doi.org/10.1080/00223131.2015.1005719

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.

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

https://doi.org/10.1016/j.net.2015.03.001

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.

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

https://doi.org/10.1155/2015/964327

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.