Calculation of hot rod heat-up during DVI LOCA accident

U radu je prikazan odziv primarnog kruga PWR elektrane (NE Krško) na akcident loma DVI (Direct Vessel Injection) cijevi ECCS (Emergency Core Cooling System) sustava. Za proračun srednjeg kanala reaktorske jezgre korišten je program RELAP5. Analizirano je stanje reaktorske posude prije i poslije UFC (Up Flow Conversion) modifikacije. Prikazane su karakteristične primarne varijable tijekom 1000 s tranzijenta. Na osnovi dobivenih vrijednosti snage, tlaka, protoka, temperature hladioca i koeficijenta prijelaza topline u jezgri proveden je proračun vruće gorive šipke korištenjem programa FRAPTRAN. Prethodno su proračunati uvjeti koji opisuju stanje goriva na kraju 29. ciklusa izgaranja NE Krško programom FRAPCON. Radijalna i aksijalna raspodjela snage u gorivu je dobivena programom PARCS. Iako je model gorive šipke u programu FRAPTRAN detaljniji, dobivene vrijednosti temperature površine košuljice su slične onima iz programa RELAP5 ako se koristi srednja linearna gustoća snage gorive šipke. Osnovna prednost provedenog proračuna je to što je osim srednje šipke moguće analizirati i bilo koju drugu, različite snage ili starosti, šipku u jezgri. Pokazano je da maksimalna temperatura košuljice goriva i trajanje otkrivanja goriva raste s porastom srednje gustoće snage i starosti goriva.The response of primary loop of PWR power plant (NPP Krsko) was shown for the case of DVI (Direct Vessel Injection) break in ECCS (Emergency Core Cooling System) line. RELAP5 code was used for calculation of core average coolant channel. The configuration of the reactor vessel was analyzed for situation before and after UFC (Up Flow Conversion) modification. The characteristic primary variables were plotted during 1000 s of transient. Using calculated values of reactor power, primary pressure, coolant temperature, and heat transfer coefficient at the surface of the fuel rod additional calculation of the hot rod was performed using FRAPTRAN code. The code needs initial conditions for the end of nuclear fuel cycle 29. Calculated using FRAPCON code. The radial and axial power distributions needed by both codes was calculated using PARCS code. The fuel rod model implemented in FRAPTRAN code is more detailed than corresponding model in RELAP5, but obtained cladding surface temperature are similar if average linear power density was used in both codes. Main benefit of performed FRAPTRAN calculation is in possibility to analyze any fuel rod from the code (having higher power or larger burnup). It was shown that maximum cladding surface temperature and duration of core uncover increases with increase of rod average power and fuel burnup