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

Narukawa, Takafumi; Kondo, Keietsu; Fujimura, Yuki; Kakiuchi, Kazuo; Udagawa, Yutaka; Nemoto, Yoshiyuki

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

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

Radio-tellurium released into the environment during the complete oxidation of fuel cladding, containment venting and reactor building failure of the Fukushima accident

Hidaka, Akihide; Kawashima, Shigeto*; Kajino, Mizuo*

Journal of Nuclear Science and Technology, 60(7), p.743 - 758, 2023/07

https://doi.org/10.1080/00223131.2022.2142311

An accurate estimation of radionuclides released during the Fukushima accident is essential. Therefore, authors investigated Te release using the Unit emission-regression estimation method, in which the deposition distribution is weighted based on the hourly deposition obtained from mesoscale meteorological model calculations assuming Unit emissions. The previous study focused on confirming the applicability of this method. Subsequent examination revealed that if any part of the time when a release have occurred is missing from the estimated release period, the entire source term calculation will be distorted. Therefore, this study performed the recalculation by extending the estimation period to cover all major releases. Consequently, unspecified release events were clarified, and their correspondence to in-core events was confirmed. The Te release caused by Zr cladding complete oxidation can explain the regional dependence of the Te/Cs ratio in the soil contamination map.

Technical basis of ECCS acceptance criteria for light-water reactors and applicability to high burnup fuel

Nagase, Fumihisa; Narukawa, Takafumi; Amaya, Masaki

JAEA-Review 2020-076, 129 Pages, 2021/03

JAEA-Review-2020-076.pdf:3.9MB

Each light-water reactor (LWR) is equipped with the Emergency Core Cooling System (ECCS) to maintain the coolability of the reactor core and to suppress the release of radioactive fission products to the environment even in a loss-of-coolant accident (LOCA) caused by breaks in the reactor coolant pressure boundary. The acceptance criteria for ECCS have been established in order to evaluate the ECCS performance and confirm the sufficient safety margin in the evaluation. The limits defined in the criteria were determined in 1975 and reviewed based on state-of-the-art knowledge in 1981. Though the fuel burnup extension and necessary improvements of cladding materials and fuel design have been conducted, the criteria have not been reviewed since then. Meanwhile, much technical knowledge has been accumulated regarding the behavior of high-burnup fuel during LOCAs and the applicability of the criteria to the high-burnup fuel. This report provides a comprehensive review of the history and technical bases of the current criteria and summarizes state-of-the-art technical findings regarding the fuel behavior during LOCAs. The applicability of the current criteria to the high-burnup fuel is also discussed.

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

Narukawa, Takafumi; Amaya, Masaki

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

Effects of ballooning and rupture on the fracture resistance of Zircaloy-4 fuel cladding tube after LOCA-simulated experiments

Yumura, Takanori; Amaya, Masaki

Annals of Nuclear Energy, 120, p.798 - 804, 2018/10

https://doi.org/10.1016/j.anucene.2018.06.046

The Effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions

Narukawa, Takafumi; Amaya, Masaki

Journal of Nuclear Science and Technology, 53(1), p.112 - 122, 2016/01

https://doi.org/10.1080/00223131.2015.1025884

Behavior of pre-hydrided Zircaloy-4 cladding under simulated LOCA conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 42(2), p.209 - 218, 2005/02

https://doi.org/10.3327/jnst.42.209

Regarding high burn-up fuel behavior under LOCA conditions, LOCA-simulated experiments were performed with unirradiated Zircaloy-4 claddings. Claddings containig 100 to 1450 ppm were isothermally oxidized at at 1220 to 1500 K in steam flow, and quenched by flooding water. Axial shrinkage of the rods during the quench was restrained controlling the maximum restraint load at four different levels. Primarily depending on fraction of the cladding thickness oxidized, the claddings fractured into two pieces during the quench, with circumferential cracking. The fracture/non-fracture threshold as for the oxidized fraction decreases as both initial hydrogen concentration and axial restraint load increase. Consequently, it was shown that the threshold is higher than 20% cladding oxidation, e.g. sufficiently higher than the limit in the Japanese ECCS acceptance criteria, irrespective of hydrogen concentration, when the restraint load is below 535 N.

Effect of pre-hydriding on thermal shock resistance of Zircaloy-4 cladding under simulated loss-of-coolant accident conditions

Nagase, Fumihisa; Fuketa, Toyoshi

Journal of Nuclear Science and Technology, 41(7), p.723 - 730, 2004/07

https://doi.org/10.3327/jnst.41.723

Experiments simulating loss-of-coolant accident (LOCA) conditions were performed to evaluate effect of pre-hydriding on thermal-shock resistance of oxidized Zircaloy-4 cladding. Artificially hydrided (400 to 600 ppm) and non-hydrided claddings were subjected to the tests. Since cladding fracture on quenching primarily depends on the oxidation amount, fracture threshold was evaluated in terms of "Equivalent Cladding Reacted (ECR)". Under axially non-restrained condition, the fracture threshold is 56% ECR and the influence of pre-hydriding is not seen. The fracture threshold is decreased by restraining the test rods on quenching, and it is more remarkable in pre-hydrided claddings than in non-hydrided claddings. Consequently, the fracture threshold was 20% ECR and 10% ECR for non-hydrided and pre-hydrided claddings, respectively, under the fully restrained condition. These results indicate possible decrease of fracture threshold of high burnup fuel claddings under LOCA conditions.

Oxidation kinetics of low-Sn Zircaloy-4 at the temperature range from 773 to 1573 K

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi

Journal of Nuclear Science and Technology, 40(4), p.213 - 219, 2003/04

https://doi.org/10.3327/jnst.40.213

Isothermal oxidation tests in flowing steam were performed on low-Sn Zircaloy-4 cladding tubes over the wide temperature range from 773 to 1573 K in order to obtain oxidation kinetics applicable to various loss-of-coolant accident conditions of LWRs. The oxidation generally obeys a parabolic rate law for the examined time range up to 3600s at temperatures from 1273 K to 1573K, and for a limited time range up to 900s from 773 to 1253 K. A cubic rate law is preferable for evaluating the longer-term oxidation at 1253 K and below. The parabolic rate law constant and the cubic rate law constant for measured weight gain were evaluated at every examined temperature, and Arrhenius-type equations were determined in order to describe the temperature dependence of the rate constants. It was indicated that the change of the oxidation kinetics from the cubic to the parabolic rate and the discontinuities in the temperature dependence of the rate constants are caused by the monoclinic/tetragonal phase transformation of ZrO.

Phenomenon identification and ranking tables (PIRTs) for loss-of-coolant accidents in pressurized and boiling water reactors containing high burnup fuel

Boyack, B. E.*; Motta, A. T.*; Peddicord, K. L.*; Alexander, C. A.*; Andersen, J. G. M.*; Blaisdell, J. A.*; Dunn, B. M.*; Ebeling-Koning, D.*; Fuketa, Toyoshi; Hache, G.*; et al.

NUREG/CR-6744, 455 Pages, 2001/12

no abstracts in English