Published 2003 | Version v1
Miscellaneous Open

A study on the mitigating capability of an auxiliary feedwater system during SBO for APR1400

  • 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)

Name Size Download all
md5:b9d48e4d0679c5addfe699a50cbfe383
188.1 kB Preview Download
Part of:
Proceedings of the 17th international conference on structural mechanics in reactor technology

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