Published October 2014 | Version v1
Miscellaneous

Pressurized Thermal Shock Analysis for OPR1000 Pressure Vessel

  • 1. Seoul National Univ., Seoul (Korea, Republic of)

Description

The study provides a brief understanding of the analysis procedure and techniques using ANSYS, such as the acceptance criteria, selection and categorization of events, thermal analysis, structural analysis including fracture mechanics assessment, crack propagation and evaluation of material properties. PTS may result from instrumentation and control malfunction, inadvertent steam dump, and postulated accidents such as smallbreak (SB) LOCA, large-break (LB) LOCA, main steam line break (MSLB), feedwater line breaks and steam generator overfill. In this study our main focus is to consider only the LB LOCA due to a cold leg break of the Optimized Power Reactor 1000 MWe (OPR1000). Consideration is given as well to the emergency core cooling system (ECCS) specific sequence with the operating parameters like pressure, temperature and time sequences. The static structural and thermal analysis to investigate the effects of PTS on RPV is the main motivation of this study. Specific surface crack effects and its propagation is also considered to measure the integrity of the RPV. This study describes the procedure for pressurized thermal shock analysis due to a loss of coolant accidental condition and emergency core cooling system operation for reactor pressure vessel.. Different accidental events that cause pressurized thermal shock to nuclear RPV that can also be analyzed in the same way. Considering the limitations of low speed computer only the static analysis is conducted. The modified LBLOCA phases and simplified geometry can is utilized to analyze the effect of PTS on RPV for general understanding not for specific specialized purpose. However, by integrating the disciplines of thermal and structural analysis, and fracture mechanics analysis a clearer understanding of the total aspect of the PTS problem has resulted. By adopting the CFD, thermal hydraulics, uncertainties and risk analysis for different type of accidental conditions, events and sequences with proper mathematical models and boundary conditions, the PTS analysis can be improved further

Part of:
Proceedings of the KNS 2014 Fall Meeting

Additional details

Publishing Information

Publisher
KNS
Imprint Place
Daejeon (Korea, Republic of)
Imprint Title
Proceedings of the KNS 2014 Fall Meeting
Imprint Pagination
[1 CD-ROM]
Journal Page Range
[7 p.]

Conference

Title
2014 Fall Meeting of the KNS
Dates
29-31 Oct 2014
Place
Pyongchang (Korea, Republic of)

INIS

Country of Publication
Korea, Republic of
Country of Input or Organization
Korea, Republic of
INIS RN
46060712
Subject category
S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
Resource subtype / Literary indicator
Conference, Non-conventional Literature
Descriptors DEI
PRESSURE VESSELS; PWR TYPE REACTORS; REACTOR CORES; REACTOR VESSELS; RISK ASSESSMENT; STEAM GENERATORS; THERMAL ANALYSIS; THERMAL SHOCK
Descriptors DEC
BOILERS; CONTAINERS; ENRICHED URANIUM REACTORS; POWER REACTORS; REACTOR COMPONENTS; REACTORS; THERMAL REACTORS; VAPOR GENERATORS; WATER COOLED REACTORS; WATER MODERATED REACTORS

Optional Information

Notes
5 refs, 23 figs, 3 tabs