PCI-related cladding failures during off-normal events - draft. [PWR; BWR]

Technical Report · Tue May 01 00:00:00 EDT 1984

OSTI ID:6767451

Van Houten, R; Tokar, M; MacDonald, P E

Pellet-cladding interaction (PCI) has long been identified as a fuel rod failure mechanism during power increases in both pressurized and boiling water reactors, and commercial guidelines have practically eliminated such failures during standard operations. A question remains regarding the possible formation of through-wall cladding cracks during several types of postulated off-normal reactor events involving power increases. This report includes preliminary findings for reactor events of the type addressed by Chapter 15 of the NRC Standard Review Plan. Specifically, the BWR turbine trip without bypass, PWR control rod withdrawal error, subcritical PWR control rod withdrawal error, BWR control blade withdrawal error, and the PWR steamline break are analyzed on the joint bases of peak rod power, power increase, ramp rate, and duration at elevated power. These Chapter 15 events are compared to numerous test reactor results and to other relevant investigations, and tentative conclusions on transient severity and data base adequacy are presented. Progress in developing computer codes for predicting PCI-induced fuel rod failures is also discussed. 49 references.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• DOE Contract Number: AC07-76ID01570
• OSTI ID: 6767451
• Report Number(s): NUREG/CR-3781-Draft; EGG-2313-Draft; ON: DE84013534
• Resource Relation: Other Information: Portions are illegible in microfiche products
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BWR TYPE REACTORS
FUEL-CLADDING INTERACTIONS
PWR TYPE REACTORS
POWER DISTRIBUTION
REACTOR ACCIDENTS
REACTOR SAFETY
TEMPERATURE GRADIENTS
THERMAL STRESSES
URANIUM DIOXIDE
ZIRCALOY
ACCIDENTS
ACTINIDE COMPOUNDS
ALLOYS
CHALCOGENIDES
OXIDES
OXYGEN COMPOUNDS
REACTORS
SAFETY
STRESSES
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM OXIDES
WATER COOLED REACTORS
WATER MODERATED REACTORS
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