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
| Identifer | oai:union.ndltd.org:UPSALLA1/oai:DiVA.org:kth-44484 |
| Date | January 2011 |
| Creators | Breijder, Paul |
| Publisher | KTH, Kärnkraftssäkerhet |
| Source Sets | DiVA Archive at Upsalla University |
| Language | English |
| Detected Language | English |
| Type | Student thesis, info:eu-repo/semantics/bachelorThesis, text |
| Format | application/pdf |
| Rights | info:eu-repo/semantics/openAccess |
| Relation | Trita-FYS, 0280-316X ; 54 |
Page generated in 0.0678 seconds