A study on the mitigating capability of an auxiliary feedwater system during SBO for APR1400
Creators
- 1. Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
Description
The objective of this paper is to establish an auxiliary feedwater (AFW) operational technical bases for the Korean Next Generation Reactor (APR1400) by modeling the plant, and by analyzing station blackout (SBO) using the MELCOR code. For the integrity of the reactor vessel and containment safety against severe accidents, it is essential to understand the severe accident sequences and to assess accident progression accurately using computer codes. Furthermore, it is important to attain the capability to analyze the advanced nuclear reactor design for the severe accident prevention and mitigation. Accident analyses are also undertaken to find out how effective AFW is mitigating in severe accident progresses. A nominal base case for SBO without AFW, time interval between feedwater stop and reactor vessel failure is 12,740 seconds. When AFW operates to mitigate the SBO accident progression 2, 4 and 8 hours after SBO starts, the reactor vessel failure is delayed for 20,415 seconds, 22,633 seconds and 26,508 seconds, respectively thus the operator has more time available for AC recovery and accident mitigation to prevent reactor vessel failure. (author)
Files
36071704.pdf
Files
(188.1 kB)
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Additional details
Additional titles
- Augmented title (English)
- [station blackout]
Publishing Information
- Publisher
- Vysoka skola technicka
- Imprint Place
- Brno (Czech Republic)
- Imprint Title
- Proceedings of the 17th international conference on structural mechanics in reactor technology
- Imprint Pagination
- [3216 p.]
- Journal Page Range
- p. 3083-3088
- Report number
- INIS-CZ--0039
Conference
- Title
- 17. international conference on structural mechanics in reactor technology
- Acronym
- SMIRT 17
- Dates
- 17-22 Aug 2003
- Place
- Prague (Czech Republic)
INIS
- Country of Publication
- Czech Republic
- Country of Input or Organization
- Czech Republic
- INIS RN
- 36071704
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- FEEDWATER; M CODES; MITIGATION; OUTAGES; PWR TYPE REACTORS; REACTOR ACCIDENTS; THERMAL HYDRAULICS
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; ENRICHED URANIUM REACTORS; FLUID MECHANICS; HYDRAULICS; HYDROGEN COMPOUNDS; MECHANICS; OXYGEN COMPOUNDS; POWER REACTORS; REACTORS; THERMAL REACTORS; WATER; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- Presented within section P01: Severe accident management and structural evaluation - Reactor pressure vessel and containment integrity. 2 tabs., 6 figs., 3 refs. Imprint:Published on CD-ROM in the PDF format for Adobe Acrobat; contact: Prof. S. Vejvoda, Institute of Applied Mechanics, Brno University of Technology, Brno, Czech Republic