未照射燃料粉末を用いた高温燃料・水間熱的相互作用の挙動評価試験

三原 武; 浦野 建太; 宇田川 豊; 垣内 一雄

JAEA-Technology 2026-009, 15 Pages, 2026/06

JAEA-Technology-2026-009.pdf:1.15MB

反応度事故(RIA)条件下で生じる照射済燃料の破損に伴って生じる機械的エネルギー(衝撃圧力や水撃力)は、燃料の細片化状態及び温度に強く依存する。低温で急速に破損するペレット/被覆管機械的相互作用(Pellet/Cladding Mechanical Interaction: PCMI)を原因とする破損では、FPガスの放出に伴ってペレット細片が水中を高速で移動し、水との超高効率な熱伝達により急激な蒸気発生が生じ、衝撃的なエネルギーが発生する。このことから、粒子表面積及びペレット細片の高速移動による極めて高い効率での熱伝達が重要な影響因子と考えられる。264-2実験及び264-24実験では、このようなRIA時に特有のペレット細片運動の駆動力が介在しない条件での燃料-水接触を模擬できるよう、試験条件を設計した。他方、264-24実験では、ペレット粒子の比表面積(単位重量あたりの表面積)を、これまで最も高かった高燃焼度燃料よりもさらに大きくなるように設計した。さらに、燃料エンタルピー(単位質量あたりの熱エネルギー)についても、従来の観測結果を踏まえ、高い機械的エネルギーの発生が期待される条件に設定した。この結果、バースト的なFPガス放出を駆動力としてペレット細片が急速に分散すると考えられる高燃焼度燃料のPCMI破損ケースと比較して、衝撃圧力及び水撃力のエネルギーのいずれも大きく下回り、機械的エネルギーの発生におけるペレット細片運動の駆動力の重要性が明瞭に示された。

Measurement of transient fission gas release from high-burnup MOX fuel under a simulated reactivity-initiated accident condition using fission gas dynamics testing technique

谷口 良徳; 浦野 建太; 三原 武; 宇田川 豊; 垣内 一雄; 勝山 仁哉

Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10

To investigate the fission gas release behavior of MOX fuel under reactivity-initiated accident (RIA) conditions, a RIA-simulated test on a high-burnup MOX fuel irradiated up to about 64.5 GWd/t (Test FGD-3) was conducted at the Nuclear Safety Research Reactor (NSRR) in JAEA by using recently developed Fission Gas Dynamics (FGD) testing technique. The concept of the FGD tests is to evaluate fission gas release during RIA-simulated test by measuring the pressure transient inside a rigid chamber containing the test fuel rod. We utilize Linear Variable Differential Transformer (LVDT)-type pressure sensor which less affected by gamma and/or neutron field in the NSRR core than conventional strain gauge-type pressure sensor. The maximum fuel enthalpy during Test FGD-3 was evaluated as 276 J/g, which is almost the same value as that of a previous FGD test on a high-burnup UO fuel (about 61 GWd/t) (Test FGD-2). The measured pressure increased from 0.1 MPa to eventually stabilized at about 0.75 MPa: this increase of pressure roughly corresponds to a transient FGR of about 28%, which is higher than that obtained in Test FGD-2 (about 18%). Sensitivity analyses of effective gas permeability for axial gas communication inside the FGD-3 test fuel rod using fuel performance code RANNS showed that apparent gas permeability of the FGD-3 fuel was much higher than that of the FGD-2 fuel. These results suggest that transient fission gas release from high-burnup MOX fuel exceeds that from UO fuel with similar burnup levels, and a significant portion released shortly after energy injection.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 三原 武; 垣内 一雄; 宇田川 豊

Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01

https://doi.org/10.1016/j.anucene.2023.110144

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5 cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.

Study on safety characteristics of a sodium-cooled fast reactor with negative void reactivity during initiating phase in severe accident

石田 真也; 深野 義隆; 飛田 吉春; 岡野 靖

Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04

One of the effective design measures against core disruptive accident (CDA) is to decrease void reactivity, and a sodium-cooled fast reactor (SFR) with low void reactivity has been developed to improve reactor safety for future SFR. The evaluation of small SFRs, which can have a negative void reactivity coefficient, is useful to examine the reactor characteristics for future research and development. The event progression of unprotected loss of flow (ULOF), which is a typical initiating event of CDA, was analyzed by the SAS4A code. In comparison with a general behaviour of large SFR with relatively higher positive void reactivity, it was confirmed that the low void reactivity reactor has the following characteristics: (1) Event progression becomes slow and mild. (2) Positive reactivity insertion by a cladding melting and relocation has larger importance. (3) Generating mechanical energy during the initiating phase becomes less likely to occur.

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; 三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

https://doi.org/10.1080/00223131.2019.1709996

To better understand the failure limit of fuel cladding during the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA), pre-cracked and hydrided cladding samples with base metal final heat-treatment status of cold worked (CW) and recrystallized (RX) were tested under biaxial stress conditions (axial to hoop strain ratios of 0 and 0.5). Displacement-controlled biaxial-expansion-due-to-compression (biaxial-EDC) tests were performed to obtain the hoop strain at failure (failure strain) of the samples. The conversion of the failure strains to J-integral at failure by finite-element analysis involving data of stress-relieved (SR) cladding specimens from our previous study revealed that the failure limit in the dimension of J-integral at failure unifies the effects of pre-crack depth. About 30 to 50 percent reduction in the J-integral at failure was observed as the strain ratio increased from 0 to 0.5 irrespective of the annealing type, pre-crack depth, and hydrogen content. the rate of fractional decreases of J-integral at failure with increase of hydrogen content are in the order of CWSRRX, which are essentially independent of strain ratio for the CW and SR samples. The results were incorporated into the failure prediction model of the JAEA's fuel performance code in the form of a correction factor that considers the biaxial loading effect.

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.

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

谷口 良徳; 宇田川 豊; 三原 武; 天谷 政樹; 垣内 一雄

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09

A pulse-irradiation test CN-1 on a high-burnup MOX fuel with M5 cladding was conducted at the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Agency (JAEA). Although the transient signals obtained during the pulse-irradiation test did not show any signs of the occurrence of PCMI failure, the failure of the test fuel rod was confirmed from the visual inspection carried out after test CN-1. Analyses using fuel performance codes FEMAXI-8 and RANNS were also performed in order to investigate the fuel behavior during normal operation and pulse-irradiation regarding the test fuel rod of CN-1, and the results were consistent with this observation result. These experimental and calculation results suggested that the failure of test fuel rod of CN-1 was not caused by hydride-assisted PCMI but high-temperature rupture following the increase in rod internal pressure. The occurrence of this failure mode might be related to the ductility remained in the M5 cladding owing to its low content of the hydrogen absorbed during normal operation.

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

天谷 政樹; 宇田川 豊; 成川 隆文; 三原 武; 谷口 良徳

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

JAEA has conducted a research program called ALPS-II program for advanced fuels of LWRs. In this program, the tests simulating a RIA and a LOCA have been performed on the high burnup advanced fuels irradiated in European commercial reactors. The failure limits of the high-burnup advanced fuels under RIA conditions have been obtained by the pulse irradiation tests at the NSRR in JAEA. The information about pellet fragmentation etc. during the pulse irradiations was also obtained from post-test examinations on the test rods after the pulse irradiation tests. As for the simulated LOCA test, integral thermal shock tests and high-temperature oxidation tests have been performed at the RFEF in JAEA. The fracture limits under LOCA and post-LOCA conditions etc. of the high-burnup advanced fuel cladding have been investigated, and it was found that in terms of these materials the fracture boundaries do not decrease and the oxidation does not significantly accelerate in the burnup level examined.

Recent research activities using NSRR on safety related issues

宇田川 豊; 杉山 智之; 天谷 政樹

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1183 - 1189, 2016/04

JAEA launched ALPS-II program in 2010 in order to obtain regulatory data for advanced fuels. Five new reactivity-initiated accident (RIA) simulated tests on the advanced fuels have been performed. The first two fuels tested, VA-5 and VA-6, were 1717-PWR-type with stress-relieved and recrystallized M-MDA cladding tube, and irradiated to ~80 GWd/tU. The cladding failed due to the pellet-cladding mechanical interaction. Fission gas dynamics tests to promote a better understanding of the behavior of fission gas during an RIA are planned. A recent qualification test on a prototype pressure sensor demonstrated its ability to obtain history data of transient fission gas release. JAEA also launched a new experiment program using NSRR to investigate fuel degradation behaviors in the temperature region beyond-DBA LOCAs.

革新的高温ガス炉燃料・黒鉛に関する技術開発計画(受託研究)

沢 和弘; 植田 祥平; 柴田 大受; 角田 淳弥; 大橋 準平; 栃尾 大輔

JAERI-Tech 2005-024, 34 Pages, 2005/03

JAERI-Tech-2005-024.pdf:2.15MB

第四世代(GEN-IV)原子炉システムの有力な候補となっている超高温ガス炉(VHTR)では、燃料は1520%FIMA,高速中性子照射量610m(E0.1MeV)においても健全性を保つ必要があるが、従来のSiC被覆燃料粒子では、このような厳しい条件下で健全性を保ったデータはない。原研で開発してきたZrC被覆燃料粒子は、SiC被覆燃料粒子よりも高温かつ高燃焼度下で健全性を維持できると期待されている。原研では、(1)従来よりも大型の被覆装置によるZrC蒸着技術の開発,(2)ZrC検査技術の開発,(3)ZrC被覆層の照射試験及び照射後試験を開始する。また、反応度投入試験を実施して被覆燃料粒子の破損機構を把握し、反応度事故時の燃料温度制限の緩和を目指す。VHTRでは、炉心の黒鉛構造物も高い中性子照射条件下で健全性を維持しなくてはならない。そこで、黒鉛構造物の超音波伝播特性や微小硬度計による圧子の押込み特性により、黒鉛構造物の機械的特性を非破壊的に評価できる技術を開発する。本報は、文部科学省からの受託研究として2004年11月から開始したこれらの研究開発の計画と2004年度の成果についてまとめたものである。