Analysis of PWR cladding transient load under LOCA quench conditions
- 1. Japan Atomic Energy Agency, Tokai, Ibaraki (Japan)
Description
LOCA is a classical design basis accident needed to be considered in all LWR safety analyses. The thermal shock induced by quench during LOCA may cause fracture in the claddings which could lead to core damage. Therefore it is necessary to study the cladding behavior during quench. This paper reports the results of LOCA quench experiments and simulations using the RANNS code for evaluating local mechanical and thermal states of axial load on the cladding. The experimental measurements suggest the rate of load gain decreases with an increasing of the ECR value due to the thicker zirconia layer which serves as a thermal barrier. In addition, the temperature-induced stress on the cladding along the axial direction appears uneven. Therefore, it is found that the LOCA simulation needs multiple elements in the axial direction for obtaining a fairly good prediction of the axial load gain. Finally, the RANNS simulation of the pellet center temperature is validated, and the RANNS code shows the capability in predicting the axial load generated during quench for ECR of lower or equal to 15%. (author)
Additional details
Publishing Information
- Imprint Title
- Proceedings of 2014 water reactor fuel performance meeting/ top fuel / LWR fuel performance meeting (WRFPM 2014)
- Imprint Pagination
- 784 p.
- Journal Page Range
- 5 p.
Conference
- Title
- 2014 water reactor fuel performance meeting/ top fuel / LWR fuel performance meeting
- Acronym
- WRFPM 2014
- Dates
- 14-17 Sep 2014
- Place
- Sendai, Miyagi (Japan)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 47070639
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- BLOWDOWN; DESIGN BASIS ACCIDENTS; ECCS; FINITE ELEMENT METHOD; HEAT TRANSFER; JAEA; LOSS OF COOLANT; QUENCHING; R CODES; REACTIVITY; REACTOR ACCIDENT SIMULATION; REACTOR CORE DISRUPTION; REACTOR SAFETY; THERMAL SHOCK; WATER COOLED REACTORS
- Descriptors DEC
- ACCIDENTS; CALCULATION METHODS; COMPUTER CODES; ENERGY TRANSFER; ENGINEERED SAFETY SYSTEMS; JAPANESE ORGANIZATIONS; MATHEMATICAL SOLUTIONS; NATIONAL ORGANIZATIONS; NUMERICAL SOLUTION; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; SAFETY; SIMULATION
Optional Information
- Notes
- Available as CD-ROM Data in PDF format, Paper ID: fullpaper100037.pdf; 12 refs., 8 figs., 1 tab.