Thermal-Hydraulic Integral Effect Test with ATLAS for an Intermediate Break Loss of Coolant Accident at a Pressurizer Surge Line
Creators
- 1. Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
Description
The main objectives of this test were not only to provide physical insight into the system response of the APR1400 during the pressurizer surge line break accident but also to produce an integral effect test data to validate the SPACE code. In order to simulate a double-ended guillotine break of a pressurizer surge line in the APR1400, the IB-SUR-01R test was performed with ATLAS. The major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Despite the core was uncovered, no excursion in the cladding temperature was observed. The pressurizer surge line break can be classified as a hot leg break from a break location point of view. Compared with a cold leg break, coolability in the core may be better in case of a hot leg break due to the enhanced flow in the core region. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for an IBLOCA simulation, especially for DVI-adapted plants. Redefinition of break size for design basis accident (DBA) based on risk information is being extensively investigated due to the potential for safety benefits and unnecessary burden reduction from current LBLOCA (large break loss of coolant accident)-based ECC (Emergency Core Cooling) Acceptance Criteria. As a transition break size (TBS), the rupture of medium-size pipe is considered to be more important than ever in risk-informed regulation (RIR)-relevant safety analysis. As plants age, are up-rated, and continue to seek improved operating efficiencies, the small break and intermediate break LOCA (IBLOCA) can become a concern. In particular, IBLOCA with DVI (Direct Vessel Injection) features will be addressed to support redefinition of a design-basis LOCA. With an aim of expanding code validation to address small and medium break sizes, development of an evaluation model (EM) for full spectrum LOCA is attracting an extensive attention in the field of nuclear safety analysis. The experimental data for an IBLOCA, however, are quite limited especially on the effect of the ECC water injection methods of DVI. Core uncovery, ECC bypass, and loop seal clearance characteristics during an IBLOCA transient may be different from those during a SBLOCA and a LBLOCA transient. In order to provide integral effect test data to validate the SPACE code for an IBLOCA, a pressurizer surge line break simulation test was decided to be performed in technical consultation with Korean nuclear industry. In this study, a thermal-hydraulic integral effect test was performed with ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) to simulate a double-ended guillotine break of a pressurizer surge line of the APR1400 (Advanced Power Reactor 1400 MWe)
Additional details
Publishing Information
- Publisher
- KNS
- Imprint Place
- Daejeon (Korea, Republic of)
- Imprint Title
- Proceedings of the KNS Fall meeting
- Imprint Pagination
- [1 CD-ROM]
- Journal Page Range
- [2 p.]
Conference
- Title
- 2013 Fall meeting of the KNS
- Dates
- 23-25 Oct 2013
- Place
- Kyungju (Korea, Republic of)
INIS
- Country of Publication
- Korea, Republic of
- Country of Input or Organization
- Korea, Republic of
- INIS RN
- 45082358
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- DESIGN; ECCS; LOSS OF COOLANT; PRESSURIZERS; PWR TYPE REACTORS; S CODES; SAFETY; SAFETY ANALYSIS; SURGES; THERMAL HYDRAULICS; VALIDATION
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; FLUID MECHANICS; HYDRAULICS; MECHANICS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; TESTING; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 4 refs, 3 figs, 1 tab