Published 2004 | Version v1
Computer medium

Best estimate approach steam line break accident analysis using uncertainty quantification method

  • 1. Safety Analysis Dept., Korea Power Engineering Company, Inc., Daejeon (Korea, Republic of)

Description

Uncertainty quantification evaluation using a best estimate methodology, the CABUE technique, was applied to the steam line break analysis for UCN 3 and 4. In the evaluation, uncertainty parameters were identified based on PIRT process and their associated distribution types and ranges were conservatively assumed. Then, the accident simulations were performed by using the RETRAN-3D code for selected uncertainty parameter sets. Reactivity margin to criticality after reactor trip was evaluate as a safety parameter. The 95 percentile of the reactivity was determined by Wilks formula from the reactivity values from 59 code simulations with 59 randomly sampled parameter vectors. The resultant reactivity was compared against that of licensing calculation. The comparison showed that applying the CABUE methodology to non-LOCA safety analysis has potential to quantify additional safety margin which was accounted for conservatively in licensing calculation. However, the success of the methodology relies on the adequacy of statistical data and identification process associated with the uncertainty parameters. In order to establish more reliable best estimate analysis methodology for non-LOCA transient analysis, more intensive and careful study is necessary on the identification of uncertainty parameters, determination of distribution types and ranges of the parameters, and computer model validation by comparison with plant transient data. (author)

Part of:
NUTHOS-6: Proceedings of the 6th international topical meeting on nuclear reactor thermal hydraulics, operations and safety

Additional details

Publishing Information

Publisher
Atomic Energy Society of Japan
Imprint Place
Tokyo (Japan)
Imprint Title
NUTHOS-6: Proceedings of the 6th international topical meeting on nuclear reactor thermal hydraulics, operations and safety
Imprint Pagination
3690 p.
Journal Page Range
11 p.

Conference

Title
6. international topical meeting on nuclear reactor thermal hydraulics, operations and safety
Acronym
NUTHOS-6
Dates
4-8 Oct 2004
Place
Nara (Japan)

INIS

Country of Publication
Japan
Country of Input or Organization
Japan
INIS RN
36093855
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
C CODES; FAILURES; NUCLEAR POWER PLANTS; PERFORMANCE; PIPES; PROBABILISTIC ESTIMATION; SAFETY ANALYSIS; STEAM; THERMAL HYDRAULICS; TRANSIENTS
Descriptors DEC
CALCULATION METHODS; COMPUTER CODES; FLUID MECHANICS; HYDRAULICS; MECHANICS; NUCLEAR FACILITIES; POWER PLANTS; THERMAL POWER PLANTS; TUBES

Optional Information

Notes
This CD-ROM can be used for WINDOWS 9x/NT/2000/ME/XP, MACINTOSH; Acrobat Reader is included; Data in PDF format, Folder Name THURSDAY October 7, 2004, Paper ID N6P 053.pdf; 13 refs., 8 figs., 1 tab.