Main feed water line break analysis for KNU 1, 9 and 10
Creators
- 1. Korea Advanced Energy Research Inst., Seoul (Republic of Korea)
Description
KNU 9,10 and KNU 1 feedwater line break transients, which are initiated by a break in feedwater piping and result in a rapid RCS heatup due to the loss of heat sink, were analyzed with RELAP 4/MOD 6. Analysis time is limited to 100 sec after break to investigate only the first pressurization transient of RCS. Comparing the results with those of SAR, several problems were discovered in SAR methods and results, however, present results also show the immediate damage of RCPB integrity not to be probable. The results show that RCS peak pressure is affected by break size greatly, so it is concluded that FLB analysis should include the sensitivity study with the parameter of break size. KNU 1 analysis results show lower RCS peak pressure than those of KNU 9, 10, so it is concluded that FKNU 1 is less dangerous than KNU 9, 10 in view of RCS pressure. And the reason was analyzed and concluded to be the difference of protection system design. (Author)
Additional details
Publishing Information
- Imprint Title
- Safety analysis on nuclear power plant
- Imprint Pagination
- 170 p.
- Journal Page Range
- pt. 4 p.59-131.
- Report number
- KAERI/RR--383-1/82
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 15025593
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- COMPUTER CALCULATIONS; FEEDWATER; HEAT SINKS; KORI-1 REACTOR; PIPES; PWR TYPE REACTORS; R CODES; REACTOR COOLING SYSTEMS; RUPTURES
- Descriptors DEC
- COMPUTER CODES; COOLING SYSTEMS; ENRICHED URANIUM REACTORS; FAILURES; HYDROGEN COMPOUNDS; OXYGEN COMPOUNDS; POWER REACTORS; REACTOR COMPONENTS; REACTORS; WATER; WATER COOLED REACTORS; WATER MODERATED REACTORS