Analysis of Advanced Fuel Behaviour during Loss of Coolant Accident in Swedish Boiling Water Reactor
Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE creditsStudent thesis
In accident analysis regarding nuclear power plants, it is very common to use thermal hydraulic system codes, such as TRACE, developed by U.S. NRC. In the case of licensing a power plant, this is one of the necessities.
TRACE is a relatively new thermal hydraulic system code and a lot of knowledge is needed to implement it in a correct way, especially in accident analysis, where it is a requirement that the rules and statements in Appendix-K, dealing with criteria for ECCS-models, are modelled. In this thesis an improved model of a Swedish Boiling Water Reactor within TRACE is realized and tested. Afterwards, once a working and representative model has been obtained, a sensitivity study in conducted in order to investigate the sensitivity of TRACE for a couple of thermal hydraulic parameters.
The sensitivity study is focussing on the eect of the peak cladding temperature, as well as the coolability of the nuclear fuel in terms of quenching and quench-front velocities.
It is found to be hard to say unilaterally what the eect of changing a certain number of parameters on the reactor behaviour is. As it turns out to be, although strongly related, the peak cladding temperatures and the quench phenomena can behave dierently
Place, publisher, year, edition, pages
2011. , 107 p.
Trita-FYS, ISSN 0280-316X ; 54
Accident analysis, BWR, LOCA, sensitivity study, thermal hydraulic system code, TRACE
IdentifiersURN: urn:nbn:se:kth:diva-44484OAI: oai:DiVA.org:kth-44484DiVA: diva2:450612
Master of Science - Nuclear Energy Engineering
UppsokPhysics, Chemistry, Mathematics