Engineering formulation of the irradiation growth behavior of zirconium-based alloys for light water reactors

垣内 一雄; 天谷 政樹; 宇田川 豊

Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01

https://doi.org/10.1016/j.jnucmat.2022.154110

The irradiation growth behavior of coupon specimens prepared from improved Zr-based alloys for light-water reactor fuel cladding, which have various additive elements and fabrication conditions, was investigated by conducting an irradiation test at 573 and 593 K under typical PWR coolant conditions up to a fast-neutron fluence of 7.810 (n/cm , E 1 MeV) in the Halden reactor in Norway. Based on the dimensional change data measured at interim and final inspections, the amounts of irradiation growth of the improved Zr-based alloys were formulated from the viewpoint of engineering. The trends of the parameters which express the effects of additive elements on irradiation growth behavior were in good agreement with those previously reported, and it was found that the amount of irradiation growth can be expressed by using a summation rule of the effect of each additive element on irradiation growth.

Mechanical property evaluation with nanoindentation method on Zircaloy-4 cladding tube after LOCA-simulated experiment

垣内 一雄; 山内 紹裕*; 天谷 政樹; 宇田川 豊; 北野 剛司*

Proceedings of TopFuel 2022 (Internet), p.409 - 418, 2022/10

In order to examine the influence of cladding microstructural changes upon the mechanical property of the fuel cladding under LOCA conditions in a more direct and quantitative manner, the nanoindentation method has been applied to Zircaloy-4 cladding specimens after LOCA simulated tests (about 1473 K, ECR 20%, quench at 973 K after slow cooling); results for two specimens taken from the rupture opening part and secondary hydriding part were compared. In addition to hardness and Young's modulus, the plastic work fraction that corresponds to the relative ductility was evaluated from the load-displacement curve. The plastic work fraction at the secondary hydriding part was found to be obviously lower than that at the rupture opening part and closer to that in -Zr(O) layers beneath the outer surface. This result from the nanoindentation method agrees with the conventional knowledge about low ductility at the secondary hydriding part.

Development of fission gas release model for MOX fuel pellets with treatment of heterogeneous microstructure

田崎 雄大; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03

https://doi.org/10.1080/00223131.2021.1973924

This study develops a new fission gas release (FGR) model for mixed oxide (MOX) fuels with a fundamentally heterogeneous microstructure. The model adopted in FEMAXI-8 was applied to irradiation Instrumented Fuel Assembly (IFA)-626 and 702 tests in which two types of MOX fuels had different heterogeneity in their microstructure, while the other spec were similar. Upon analyzing these fuels, the original FGR model predicted lower FGR from the fuel with a remarkably heterogeneous microstructure than the other MOX fuel. This estimation contradicts the experimental observation. However, the new FGR model improved the consistency because of the early release of fission gas from Pu agglomerate region, and showed issues for aiming further improvement. Therefore, the above results confirmed a certain validity of the developed model for studying heterogeneity effect.

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.

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions; Results of BZ-3 and BZ-4 tests

垣内 一雄; 宇田川 豊; 天谷 政樹

Annals of Nuclear Energy, 155, p.108171_1 - 108171_11, 2021/06

https://doi.org/10.1016/j.anucene.2021.108171

In order to investigate fission gas release behavior of high-burnup mixed-oxide (MOX) fuel pellet for LWR under reactivity-initiated accident (RIA), the tests called BZ-3 and BZ-4 were conducted at the Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Agency (JAEA). Electron probe microanalysis and rod-puncture tests were performed on the fuel pellets before and after pulse irradiation tests, and from the comparison between the puncture test results and the results evaluated from EPMA, it was suggested that fission gas release from not only the Pu-spot but also the Pu-spot-excluded region.

Evaluation of the maximum bending stress of pre-hydrided Zircaloy-4 cladding tube after simulated loss-of-coolant-accident test

岡田 裕史; 天谷 政樹

Annals of Nuclear Energy, 145, p.107539_1 - 107539_8, 2020/09

https://doi.org/10.1016/j.anucene.2020.107539

In order to evaluate the fracture resistance of fuel rods against a seismic loading which might be applied following a Loss-Of-Coolant-Accident (LOCA), the bending strength of fuel cladding which experienced a simulated LOCA has been investigated after the Fukushima-Daiichi Nuclear Power Plant accident. In consideration of previous studies and results, the effect of the amount of oxidation on the maximum bending stress of pre-hydrided cladding tube with a small amount of ballooning was investigated in this study. According to the obtained results, it was suggested that the decrease in the maximum bending stress of the cladding tube experienced LOCA conditions is mainly determined by the hydrogen concentration in the cladding tube after simulated LOCA test, irrespective of pre-hydriding. It was also suggested that the decreasing trend of the maximum bending stress with increasing the hydrogen concentration would be expressed by a form of exponential function, in which the maximum bending stress at a hydrogen concentration of 1500 ppm was estimated to be about a half of that at 0 ppm.

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.

The Effect of base irradiation on failure behaviors of UO and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

宇田川 豊; 三原 武; 谷口 良徳; 垣内 一雄; 天谷 政樹

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

AA2019-0372.pdf:0.81MB

https://doi.org/10.1016/j.anucene.2019.107268

This paper 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 cladding hoop-stress more than 50 MPa discriminates the OS-1 rod from other BWR rods and 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.

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

宇田川 豊; 杉山 智之; 天谷 政樹

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

https://doi.org/10.1080/00223131.2019.1637795

反応度事故時のペレット・被覆管相互作用により生じる軽水炉燃料の破損に関して、我が国の規制基準改訂の検討に資するため、原子炉安全性研究炉NSRRを用いて得られた近年の研究成果を総括する。これに基づき、現行基準の妥当性及び現行基準に代わりうる新たな判断基準としての燃料破損しきい値とその考え方について議論する。

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.