Failure-Bearing Capability of Oxidized Zircaloy-4 Cladding under Simulated Loss-of-Coolant Condition

LOCA-simulated experiments were performed with MDA, ZIRLOÔ, M5 Ò , NDA, and Zircaloy-2 cladding specimens with local burn-ups ranging from 66 to 76 MWd/kg. Short test rods fabricated with the cladding specimens were heated, isothermally oxidized at 1,459 to 1,480 K in steam flow, and finally quenched in flooding water. Rod rupture and subsequent double-sided oxidation of the cladding were also simulated in the experiments. Neither split-fracture nor fragmentation occurred during the quench in the cladding specimens which were oxidized to about 18-27% of the metallic thickness. Accordingly, the fracture boundary, a most important safety issue, is not reduced significantly by the high burn-up and use of the new alloys within the examined scope, although it may be somewhat reduced with pre-hydriding during the reactor operation as observed in unirradiated specimens.

Section: Resultsmentioning

confidence: 99%

“…As discussed in previous reports, 2,4,5) axial shrinkage of the fuel rods during the cooling and quench may be restrained in the bundle geometry. Hence, the test rods have been cooled and quenched under the restrained conditions in the experiments.…”

Section: Apparatus and Test Proceduresmentioning

confidence: 95%

Behavior of High Burn-up Fuel Cladding under LOCA Conditions

Nagase

Chuto

Fuketa

2009

“…In the ruptured cladding, circumferential increase, rupture opening, oxidation, and hydrogen pickup are observed in the axial length of about 60 mm [1][2][3]5]. Therefore, it is difficult to evaluate the mechanical properties of the ruptured cladding with small specimens, such as ring-like specimens, with lengths of approximately 10 mm.…”

Section: Introductionmentioning

confidence: 96%

“…The acceptance criteria for the ECCS define a maximum allowable amount of oxidation (15%) and peak cladding temperature (1200 • C) in order to avoid excessive embrittlement of the fuel cladding and to maintain the coolable geometry of a reactor core during a LOCA. These criteria were determined based on thermal shock resistance data with the cladding that was ruptured, oxidized, and quenched under the simulated LOCA conditions in Japan [1][2][3]. It has been assumed that the greatest impact on fuel rods is a thermal shock that occurs during a quench.…”

Section: Introductionmentioning

confidence: 99%

Evaluation of fracture resistance of ruptured, oxidized, and quenched Zircaloy cladding by four-point-bend tests

Yamato

Nagase

Amaya

2014

To evaluate fracture resistance of light water reactor fuel rods under loss-of-coolant accident (LOCA) and post-LOCA cooling conditions, four-point-bend tests were performed on non-irradiated Zircaloy cladding samples that were ruptured, oxidized in high-temperature steam, and quenched in flooding water. The bend test methodology was designed to apply a uniform bending moment to the entire rupture region and to generate tensile stress on the ruptured side. The fracture bending moment of the cladding decreased with oxidation temperature and hydrogen concentration as well as oxidation amount. Comparison with bending moments estimated from design basis seismic ground motion indicated that the cladding is unlikely to be fractured by seismic loads during post-LOCA cooling if high-temperature oxidation is kept below 15% equivalent cladding reacted, the oxidation limit of the Japanese LOCA criteria.

“…The temperature-time boundary separating these two categories defines the so called LOCA zero ductility limit 8) and is normally expressed by the extent of cladding oxidation calculated at that boundary by means of the Baker-Just correlation (ECR BJ ), as done for instance in. 9) In other countries the LOCA criteria are based on integral quench tests conducted on cladding tube segments. These tests have been performed at e.g.…”

Section: Loca Limit 1 Basis For the Loca Criteriamentioning

confidence: 99%

RIA Failure Threshold and LOCA Limit at High Burn-up

Vitanza

2006

A correlation for RIA fuel failure threshold has been derived and compared with recent experimental data. The correlation can be used for UO 2 and MOX fuel and at hot zero power and cold zero power transients, account taken for the different initial enthalpy and for the lower ductility at cold conditions. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. A probabilistic approach has been attempted for the cold zero power case. For LOCA, a correlation has been derived expressing the LOCA limit as a function of the hydrogen accumulated in the cladding during base irradiation. However, LOCA experimental data exhibit a significant scatter, possibly because results depend on details in the conduct of the tests. Consequently, predictions of LOCA limits vs. burn-up are affected by appreciable uncertainty, and more data derived from testing of high burn-up cladding specimens are needed in order to reach firm conclusions.