OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Influence of pellet geometry and gap size

Soba, A.*; Prudil, A.*; Zhang, J.*; Dethioux, A.*; Han, Z.*; Dostal, M.*; Matocha, V.*; Marelle, V.*; Lasnel-Payan, J.*; Kulacsy, K.*; et al.

Proceedings of TopFuel 2021 (Internet), 10 Pages, 2021/10

The NEA Expert Group on Reactor Fuel Performance (EGRFP) proposed a benchmark on fuel performance codes modeling of pellet-cladding mechanical interation (PCMI). The aim of the benchmark was to improve understanding and modeling of PCMI amongst NEA member organizations. This was achieved by comparing PCMI predictions for a number of specified cases. The results of the two hypothetical cases (1 and 2) were presented earlier. The two final cases (3 and 4) are comparison between calculations and measurements, which will be published as NEA reports. This paper focuses on Case 3, which consists of eight beginning of life (BOL) sub-cases (3a to 3h) each with different pellet designs that have undergone ramping in the Halden Reactor. The aforementioned experiments are known as the IFA-118 experiments and were performed from 1969 to 1970. The variations between cases include four different pellets dimensions (7, 14, 20 and 30 mm of height), two different gapsizes between pellet-cladding (40 and 100 microns) and three variations on pellet face geometry (flat, dishing and dishing with chamfer). Such diversity has allowed exploring the codes sensitivity to these individual factors.

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

谷口 良徳; 宇田川 豊; 天谷 政樹

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

https://doi.org/10.1016/j.anucene.2019.107188

The fuel-failure-limit data obtained in the simulated reactivity-initiated-accident experiment SPERT-CDC 859 (SPERT859) has entailed a lot of discussions if it represents fuel-failure behavior of typical commercial LWRs for its specific pre-irradiation condition and fuel state. The fuel-rod conditions before and during SPERT859 were thus assessed by the fuel-performance codes FEMAXI-8 and RANNS with focusing on cladding corrosion and its effect on the failure limit of the test rod. The analysis showed that the fuel cladding was probably excessively corroded even when the influential calculation conditions such as fuel swelling and creep models were determined so that the lowest limit of the cladding oxide layer thickness was captured. Such assumption of excessive cladding corrosion during pre-irradiation well explains not only the test-rod state before pulse irradiation but also the fuel-failure limit observed. Such understanding undermines anew the representativeness of the test data as a direct basis of safety evaluation for LWR fuels.

OECD/NEA benchmark on pellet-clad mechanical interaction modelling with fuel performance codes; Impact of number of radial pellet cracks and pellet-clad friction coefficient

Dostl, M.*; Rossiter, G.*; Dethioux, A.*; Zhang, J.*; 天谷 政樹; Rozzia, D.*; Williamson, R.*; Kozlowski, T.*; Hill, I.*; Martin, J.-F.*

Proceedings of Annual Topical Meeting on Reactor Fuel Performance (TopFuel 2018) (Internet), 10 Pages, 2018/10

The benchmark on PCMI was initiated by OECD/NEA Expert Group on Reactor Fuel Performance (EGRFP) in June 2015 and is currently in the latter stages of compiling results and preparing the final report. The aim of the benchmark is to improve understanding and modelling of PCMI amongst NEA member organisations. This is being achieved by comparing PCMI predictions of different fuel performance codes for a number of cases. Two of these cases are hypothetical cases aiming to facilitate understanding of the effects of code-to-code differences in fuel performance models. The two remaining cases are actual irradiations, where code predictions are compared with measured data. During analysis of participants' results of the hypothetical cases, the assumptions for number of radial pellet cracks and the pellet-clad friction coefficient (which can be zero, finite or infinite) were identified to be important factors in explaining differences between predictions once pellet-cladding contact occurs. However, these parameters varied in the models and codes used originally by the participants. This fact led to the extension of the benchmark by inclusion of two additional cases, where the number of radial pellet cracks and three different values of the friction coefficient were prescribed in the case definition. Seven calculations from six organisations contributed results were compared and analysed in this paper.

Pellet-cladding mechanical interaction of PWR fuel rod under rapid power transient

柳澤 和章; 片西 昌司; 藤城 俊夫

Journal of Nuclear Science and Technology, 31(7), p.671 - 676, 1994/07

https://doi.org/10.3327/jnst.31.671

出力過渡実験(燃料棒出力を4kW/mから32kW/mまで3kW/msで急昇)を実施し、以下の知見を得た。長手方向の燃料棒伸び歪みで評価したペレット-被覆管の力学的相互作用(PCMI)は、ベース照射済PWR燃料で0.07%であり、未照射PWR燃料で0.08%であった。観察された歪みの大きさは、スウェーデンのTRANS RAMPII実験での破損燃料の歪みと同程度であった。それにも拘らず本実験で燃料破損が生じなかった主たる理由は、線出力密度の増加途中から著しくなった緩和によるPCMIの急激な低下が原因であったと判断される。PWR燃料棒に発生したPCMIが、出力保持中ではなく過渡出力増加中に緩和によって減少する現象は本実験を通じて初めてみつけられた事象であり、燃料破損防止対策上極めて有用なものである。

軽水炉燃料のペレット-被覆相互作用, (I); PWR型燃料棒

柳澤 和章; 近藤 吉明*; E.Kolstad*

日本原子力学会誌, 28(7), p.641 - 657, 1986/00

https://doi.org/10.3327/jaesj.28.641

1717PWR型燃料棒のペレット-被服相互作用(PCI)に及ぼす燃焼度と燃料設計因子の効果を研究した。その結果、(1)燃焼度の増加はPCIを加速すること、(2)直径ギャップを小さくしたり、加圧量を小さくすると、PCIが加速されること、また両端ディッシュに両面取りをかけ、更にL/Dを小さくすると、PCIが低下すること、が明らかになった。(3)燃焼度18MWd/kgUで、周期20~25分の出力サイクル実験を、線出力45kW/mと30kW/mの間で、620回実施したが、燃料は健全であった。サイクル中に5~10%の範囲で出力オーバーシュートが生じ、径方向PCIはその都度、約0.05%増加した。出力サイクル後の燃料棒内圧値は、サイクル棒とレファレンス棒ほぼ同じで、約3.80.1MPaであった。即ち、サイクリングによるFPガス放出率の加速傾向は殆んどなかった。燃料からのFPガス放出率は、拡散律速の√t(tは時間)依存型であった。(4)実験中、1本の燃料棒にPCI破損が生じたが、それは燃焼度12MWd/kgU,線出力50kW/m,直径歪0.25%,ガス増加圧力2.0MPa等の条件下にあった。

NSRR実験プログレス・レポート,8; 1979年1月~1979年6月

石川 迪夫

JAERI-M 8779, 75 Pages, 1980/03

JAERI-M-8779.pdf:2.86MB

本報告書は、1979年1月から同年6月までにNSRRにおいて実施した燃料破損実験の結果およびその考察等についてまとめたものである。今期実施した実験は、燃料パラメータ試験(特殊被覆材試験、ギャップガスパラメー夕試験)、冷却材パラメータ試験(強制対流試験、低サブクール強制対流試験)、USNRC燃料試験、欠陥燃料試験(浸水燃料試験)、特殊燃料試験(燃料メルトダウン試験)および高温高圧力プセル試験(特性測定および発熱量較正試験)の総計47回である。

Comparative analysis on base-irradiation behaviors of OS-1 test rod and other BWR-fuel rods subjected to previous NSRR tests

宇田川 豊

no journal, , 

This presentation reports a computer-code analysis on the base-irradiation behavior of the chromia-and-alumina-doped BWR rod irradiated to 64 GWd/t in Oskarshamn-3, Sweden, and subjected to the reactivity-initiated-accident (RIA) test OS-1, which resulted in a fuel failure due to pellet-cladding mechanical interaction (PCMI) at the lowest fuel-enthalpy increase in all the BWR tests ever performed. The inverse calculation which utilized post-irradiation examination data as its constraint conditions revealed that the OS-1 rod had very likely experienced more intense PCMI loading due to higher swelling rate during base irradiation than other BWR rods subjected to previous RIA tests and thus had been prone to experience enhanced radial-hydride formation. The significant difference in the PCMI-related parameters between the OS-1 rod and other BWR rods supports the interpretation that enhanced radial-hydrides formation differentiated the PCMI-failure behavior observed in the test OS-1 from the previous BWR-fuel tests.