Published 1980 | Version v1
Report Open

Assessment of light water reactor fuel damage during a reactivity initiated accident

Creators

Description

This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalpy of approximately 140 cal/g UO2. Volume expansion of previously irradiated fuel upon melting may cause deformation and rupture of the cladding, and coolant channel blockage at higher peak enthalpies

Availability note (English)

MF available from INIS under the Report Number; Available from NTIS., PC A04/MF A01.

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Additional details

Publishing Information

Imprint Pagination
60 p.
Report number
CONF-800971--1

Conference

Title
CNSI specialist meeting on safety aspects of fuel behavior in off-normal and accident conditions.
Dates
1 - 4 Sep 1980.
Place
Helsinki, Finland.

INIS

Country of Publication
United States
Country of Input or Organization
United States
INIS RN
12581572
Subject category
S61: RADIATION PROTECTION AND DOSIMETRY; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
BWR TYPE REACTORS; DEFORMATION; FUEL CANS; FUEL RODS; MICROSTRUCTURE; PWR TYPE REACTORS; REACTIVITY INSERTIONS; REACTOR ACCIDENTS; REACTOR SAFETY; SPATIAL DISTRIBUTION; STRAINS; THERMAL STRESSES; ZIRCALOY
Descriptors DEC
ACCIDENTS; ALLOYS; CRYSTAL STRUCTURE; DISTRIBUTION; FUEL ELEMENTS; REACTOR COMPONENTS; REACTORS; SAFETY; STRESSES; TIN ALLOYS; TRANSITION ELEMENT ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS