Nuclear fuel performance evaluation. Final report. [POSHO code]

Technical Report · Wed Jun 01 04:00:00 EDT 1977

DOI:https://doi.org/10.2172/7089013· OSTI ID:7089013

Boerresen, S.; Pomeroy, D. L.; Rolstad, E.; Sauar, T. O.

An evaluation has been made of the ability of Scandpower's empirical fuel performance model POSHO (''Power Shock'') to predict the probability of fuel pin failures resulting from pellet-clad interaction in commercial nuclear power plants. POSHO provides an analytical method to calculate the failure probabilities associated with power level maneuvers for different fuel assembly designs. Application of the method provides a basis for risk-benefit decisions concerning operational procedures, fuel designs and fuel management strategies. One boiling water reactor (BWR) and one pressurized water reactor (PWR) were selected for study to compare model predictions with actual failures, as determined from post irradiation examination of the fuel and activity release data. The fuel duty cycles were reconstructed from operating records and nodal power histories were created by using Scandpower's FMS computer programs. Nodal power histories, coupled with the relative pin power distribution in each node, were processed by the fuel failure prediction model, which tracks the interaction power level for each pin group in each node and calculates the power shocks and the probability for pellet-clad interaction cracks. The results of these calculations are processed statistically to give the expected number of cracks, the number of failed fuel pins in each assembly and the total number of failed assemblies in the core. Fuel performance in the BWR, Quad Cities Unit Two, was calculated by the model in approximate agreement with the observed performance. Fuel performance in the PWR, Maine Yankee, was calculated in approximate agreement for two of the three fuel designs. The high failure rate in the third design, Type B fuel, was not calculated by the POSHO pellet-clad interaction model.

• Research Organization: Scandpower, Inc., Bethesda, Md. (USA)
• OSTI ID: 7089013
• Report Number(s): EPRI-NP-409(6-77)
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
210200 -- Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
ACCIDENTS
BWR TYPE REACTORS
COMPUTER CODES
ENRICHED URANIUM REACTORS
FUEL ELEMENT FAILURE
FUEL-CLADDING INTERACTIONS
MAINE YANKEE REACTOR
P CODES
POWER DISTRIBUTION
POWER REACTORS
PROBABILITY
PWR TYPE REACTORS
QUAD CITIES-2 REACTOR
REACTOR ACCIDENTS
REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS