Safety margins in zircaloy oxidation and embrittlement criteria for emergency core cooling system acceptance

Journal Article · Mon Sep 01 00:00:00 EDT 1986 · Nucl. Technol.; (United States)

OSTI ID:5121525

Williford, R E

Current emergency core cooling system acceptance criteria for light water reactors specify that, under loss-of-coolant accident (LOCA) conditions, the Baker-Just (BJ) correlation must be used to calculate Zircaloy-steam oxidation, calculated peak cladding temperatures (PCT) must not exceed 1204/sup 0/C, and calculated oxidation must not exceed 17% equivalent cladding reacted (ECR). An appropriately defined minimum margin of safety was estimated for each of these criteria. The currently required BJ oxidation correlation provides margins only over the 1100 to 1500/sup 0/C temperature range at the 95% confidence level. The PCT margins for thermal shock and handling failures are adequate at oxidation temperatures above 1204/sup 0/C for up to 210 and 160 s, respectively, at the 95% confidence level. The ECR thermal shock and handling margins at the 50 and 95% confidence levels, respectively, range between 2 and 7% ECR for the BJ correlation, but vanish at temperatures above 1100 to 1160/sup 0/C for the best-estimate Cathcart-Pawel correlation. However, use of the Cathcart Pawel correlation for ''design basis'' LOCA calculations can be justified at the 85 to 88% confidence level if cooling rate effects can be neglected.

• Research Organization: Pacific Northwest Lab., Physical Sciences Building, Battelle Boulevard, Richland, WA 99352
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 5121525
• Journal Information: Nucl. Technol.; (United States), Vol. 74:3
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
ECCS
MATERIALS TESTING
TEMPERATURE EFFECTS
WATER COOLED REACTORS
LOSS OF COOLANT
ZIRCALOY
EMBRITTLEMENT
OXIDATION
COOLING
CORRELATIONS
DESIGN BASIS ACCIDENTS
FAILURES
FUEL-CLADDING INTERACTIONS
PERFORMANCE TESTING
REACTOR SAFETY
STEAM
THERMAL SHOCK
VERY HIGH TEMPERATURE
ACCIDENTS
ALLOYS
CHEMICAL REACTIONS
ENGINEERED SAFETY SYSTEMS
REACTOR ACCIDENTS
REACTOR PROTECTION SYSTEMS
REACTORS
SAFETY
TESTING
TIN ALLOYS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
220900* - Nuclear Reactor Technology- Reactor Safety
210100 - Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
360105 - Metals & Alloys- Corrosion & Erosion
360103 - Metals & Alloys- Mechanical Properties