Oxidation and embrittlement behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之

Journal of Nuclear Materials, 587, p.154736_1 - 154736_8, 2023/12

https://doi.org/10.1016/j.jnucmat.2023.154736

To evaluate the oxidation and embrittlement behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions, we conducted isothermal oxidation and ring-compression tests on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens. Further, we discussed the loss of coolable geometry of the reactor core loaded with the FeCrAl-ODS cladding tubes under LOCA conditions, using data from the ring-compression tests in this study and the integral thermal shock tests from our previous study. The results reveal that oxidation kinetics of the FeCrAl-ODS cladding tube at 1523 K is four orders of magnitude lower than that of a conventional Zircaloy cladding tube, which highlights the exceptional oxidation resistance of the FeCrAl-ODS cladding tube. The breakaway oxidation of the FeCrAl-ODS cladding tube was observed at 1623 K for durations equal to or exceeding 6 h, and melting was observed at 1723 K. The ring-compression and the integral thermal shock tests indicate that, depending on the oxidation time, the ductile to brittle transition threshold - as determined by the ring-compression test - exists between 1623 K and 1723 K. Meanwhile, the fracture threshold - established through the integral thermal shock test - falls between 1573 K and 1673 K. Therefore, taking a conservative approach based on available data, the fracture and non-fracture results from the integral thermal shock tests can define the lower and upper boundaries of the threshold for the loss of coolable geometry of the reactor core during a LOCA.

Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

成川 隆文; 近藤 啓悦; 藤村 由希; 垣内 一雄; 宇田川 豊; 根本 義之

Journal of Nuclear Materials, 582, p.154467_1 - 154467_12, 2023/08

https://doi.org/10.1016/j.jnucmat.2023.154467

To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200-300 K higher than that of the Zircaloy cladding tube, and the FeCrAl-ODS cladding tube's maximum circumferential strain was smaller than or equal to the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of 5000 N, which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at 1473 K for 1 h. The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at 1673 K and fractured after abnormal oxidation at 1573 K for 1 h. Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below 1473 K, where no melting or abnormal oxidation occurs.

Hierarchical Bayes model to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under LOCA conditions

成川 隆文; 濱口 修輔*; 高田 孝*; 宇田川 豊

Proceedings of Asian Symposium on Risk Assessment and Management 2022 (ASRAM 2022) (Internet), 11 Pages, 2022/12

To realize a more reliable safety evaluation of loss-of-coolant accidents (LOCAs) in light-water-reactors, we developed a quantification method of the fracture limit uncertainty of high-burnup advanced fuel cladding tubes using a hierarchical Bayes model that can quantify uncertainty even when experimental data are limited. The fracture limit uncertainty was quantified as a probability using the amount of oxidation and the initial hydrogen concentration (the hydrogen concentration in fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. The hierarchical Bayes model was developed by dividing the regression coefficients into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences between types of fuel cladding tubes. Using the developed model, we showed that the fracture limits of the high-burnup advanced fuel cladding tubes tended to be on average equal to or higher than that of an unirradiated conventional fuel cladding tube. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data depending on the condition of the fuel cladding tube specimens after the LOCA-simulated test instead of the binary data, thereby increasing the amount of information in each data.

Four-point-bend tests on high-burnup advanced fuel cladding tubes after exposure to simulated LOCA conditions

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 57(7), p.782 - 791, 2020/07

https://doi.org/10.1080/00223131.2020.1724207

To evaluate the fracture resistance of high-burnup advanced fuel cladding tubes during the long-term core cooling period following loss-of-coolant accidents (LOCAs), laboratory-scale four-point-bend tests were performed using the following advanced fuel cladding tubes with burnups of 73 - 84 GWd/t: low-tin ZIRLO, M5, and Zircaloy-2 (LK3). Three four-point-bend tests were performed on the high-burnup advanced fuel cladding tube specimens subjected to the integral thermal shock tests which simulated LOCA conditions (ballooning and rupture, oxidation in high-temperature steam, and quench). During the four-point-bend tests, all the specimens that were oxidized at 1474 K to 9.9% - 21.5% equivalent cladding reacted exhibited brittle fractures. The maximum bending moments were comparable to those of the conventional Zircaloy cladding tube specimens. Furthermore, the effects of oxidation and hydriding on the maximum bending moment were comparable between the high-burnup advanced fuel cladding tube specimens and the unirradiated Zircaloy-4 cladding tube specimens. Therefore, it can be concluded that the post-LOCA fracture resistance of fuel cladding tubes is not significantly reduced by extending the burnup to 84 GWd/t and using the advanced fuel cladding tubes, though it may slightly decrease with increasing initial hydrogen concentration in a relatively lower ECR range ( 15%), as observed for the unirradiated Zircaloy-4 cladding tubes.

Behavior of high-burnup advanced LWR fuel cladding tubes under LOCA conditions

成川 隆文; 天谷 政樹

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.912 - 921, 2019/09

To evaluate behavior of high-burnup advanced light-water-reactor fuel cladding tubes under loss-of-coolant accident conditions, laboratory-scale isothermal oxidation tests and integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73-85 GWd/t: M-MDA, low-tin ZIRLO, M5textregistered, and Zircaloy-2 (LK3). The isothermal oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube and was slower than that given by the Baker-Just oxidation rate equation. Therefore, the oxidation kinetics is considered to be not significantly accelerated by extending the burnup and changing the alloy composition. During the integral thermal shock tests, the high-burnup advanced fuel cladding tube specimens did not fracture under the oxidation condition equivalent to or lower than the fracture limit of the unirradiated Zircaloy-4 cladding tube. Therefore, the fracture limit of fuel cladding tubes is considered to be not significantly reduced by extending the burnup and changing the alloy composition, though it may slightly decrease with increasing initial hydrogen concentration.

EDC試験手法による反応度事故時の燃料被覆管破損に及ぼす水素化物偏在及び2軸応力状態の影響の評価

篠崎 崇; 三原 武; 宇田川 豊; 杉山 智之; 天谷 政樹

JAEA-Research 2014-025, 34 Pages, 2014/12

JAEA-Research-2014-025.pdf:6.05MB

EDC(Expansion-Due-to-Compression)試験は、燃料被覆管の機械特性試験の一手法であり、反応度事故(RIA)時におけるペレット-被覆管機械的相互作用(PCMI)に着目した試験手法である。本研究では、高燃焼度燃料被覆管に見られる"水素化物リム"を模擬するために外周部に水素化物を偏析させた未照射被覆管を使用し、高燃焼度燃料のRIA時に被覆管に負荷される機械的条件を模擬したEDC試験を実施した。試料の水素濃度および偏析した水素化物の厚みが増加すると、試験後試料の周方向残留ひずみが低下する傾向が見られた。また、RIA時に被覆管外面の水素化物に発生するき裂を模擬するため、外面に予き裂を有する被覆管(RAG管)を作製し、この試料を対象としたEDC試験を行った結果、試料の予き裂深さが増加するにつれて破損時の周方向全ひずみが低下する傾向が見られた。さらに、RAG管試料に軸方向引張荷重を負荷することで2軸応力状態とし、EDC試験を実施した。このような2軸応力状態では、単軸引張条件である通常のEDC試験と比較して破損時の周方向全ひずみが低下する傾向が見られた。

燃料被覆管の熱変形挙動評価試験技術の開発(受託研究)

金子 哲治; 塚谷 一郎; 木内 清

JAERI-Tech 2004-035, 18 Pages, 2004/03

JAERI-Tech-2004-035.pdf:0.81MB

低減速軽水炉用燃料は、高転換比と高燃焼度化を同時に達成するために、MOX燃料とUOブランケットの各ペレット燃料域の積層構造を有している。当該燃料棒は、現用ABWR燃料と比較して、長手方向における不均一な線出力密度分布に伴う熱応力が加わることが特徴である。そのためMOX燃料とUOブランケットに起因した異なる温度分布を持った被覆管の局所的変形挙動の評価が最も重要となる。そのような力学的特性評価試験法として、短尺の被覆管試験片を用いて、実用条件で想定される当該燃料棒の一段の積層部における2軸応力下での熱疲労挙動が再現できる力学的特性評価試験装置を設計した。本装置は、温度分布制御用加熱部,軸方向疲労要素負荷用低サイクル疲労制御部及び内圧疲労要素用の内圧負荷部から構成され、局所的な変形挙動が高精度で測定できる。また、本装置により、炉の起動停止や制御等の運転モードが関係した負荷変動,燃料棒の拘束条件,燃焼度に伴うFP内圧変化の試験を行うことが可能である。

冷却材喪失事故後の燃料被覆管耐破断性評価

成川 隆文; 宇田川 豊; 天谷 政樹

no journal, , 

冷却材喪失事故(LOCA)後に長期にわたって炉心の冷却可能形状を維持できるか否かを評価するために必要な、燃料被覆管のLOCA後耐破断性に及ぼすLOCA時二次水素化の影響、並びに燃焼の進展及び被覆管材質変更の影響を評価することを目的に、LOCA模擬試験後の非照射燃料被覆管及び高燃焼度改良型燃料被覆管に対し4点曲げ試験を実施し、作用する曲げに対する耐破断性を評価した。その結果、燃料被覆管の二次水素化した領域は酸化のみの領域(破裂開口部)に比べ最大曲げ応力が半分程度に低下し、LOCA時の膨れ量及び酸化量次第で燃料被覆管は二次水素化部で破断する可能性があること、並びにLOCA後耐破断性は通常運転中の水素吸収量の増大に伴い若干低下するものの約85GWd/tまでの燃焼度進展や被覆管材質の変更により著しく低下しないことを明らかにした。