Zircaloy Fuel Cladding Behavior in a Loss-of-Coolant Accident: A Review

Abstract

This paper reviews the state-of-the-art experimental work performed in several countries with respect to the acceptance criteria established for the emergency core cooling (ECC) in a loss-of-coolant accident (LOCA) of light water reactors (LWRs). It covers in detail oxidation, embrittlement, plastic deformation, and coolability of deformed rod bundles. The main test results are discussed on the basis of research work performed at the Karlsruhe Nuclear Research Center (KfK) within the framework of the Nuclear Safety Project (PNS). Reference is made to test data obtained in other countries. The paper concludes that the major mechanisms and consequences of oxidation, deformation, and emergency core cooling are sufficiently investigated in order to provide a reliable data base for safety assessments and licensing of LWRs. All test data prove that the ECC criteria are conservative and that the coolability of a LWR and the public safety in a LOCA can be maintained.

Study of the response of Zircaloy- 4 cladding to thermal shock during water quenching after double sided steam oxidation at elevated temperatures

“…This criterion is used in the Canadian ELOCA licensing code for PHWR CANDU fuel. This criterion is also consistent with international practice where a number of acceptance criteria for preventing fuel failure due to oxygen and hydrogen embrittlement with water quench by the emergency core cooling system have been proposed by Furuta et al [26], Erbacher et al [27], Chung [28], and Nagase et al [29]. The results of the present investigation shows the thermal shock boundary to be 1050 C, 29% ECR, 290 mm of combined thickness of …”

The use of micro-Raman imaging to measure 18 O tracer distribution in thermally grown zirconia scales

“…Corrosion of cladding materials has been the subject of numerous studies since decades [1]. Cladding behavior at high temperature in conditions representative of accidental situations is also an important concern [2].…”

Modelling and Simulation of Reactor Fuel Cladding under Loss-of-Coolant Accident Conditions

“…More specifically, the azimuthal temperature variations (ÁT) of the tests reported by Erbacher et al, 4) Erbacher and Leistikow, 28) and Karb et al 27) were ÁT < 15 K, 20 < ÁT < 70 K, and 0 < ÁT < 80 K, respectively. The burst test data provided by Erbacher et al 4) are the most appropriate among the three data sets for our comparison, since the influence of azimuthal temperature variation is least for this data set (ÁT < 15 K).…”

“…The heating rate of the unirradiated rods varied between 7 to 19 K s À1 , whereas that of the electrically heated rod simulators had heating rates of 12-13 K s À1 . Erbacher and Leistikow 28) have presented Zircaloy-4 burst data obtained from multirod burst tests performed within the REBEKA program. The data represent tests that had the potential for maximum ballooning, i.e., burst taking place in the -phase of Zircaloy around 800 C. The heating rate during heat-up in the tests was 7 K s azimuthal temperature difference around the circumference of the cladding tube.…”