Assessment of the CATHARE 3-D module for LBLOCA simulation
Creators
- 1. Commissariat a l'Energie Atomique, CEA - GRENOBLE, 17 rue des Martyrs, 38054 GRENOBLE cedex 9 (France)
Description
Full text of publication follows: CATHARE is a best-estimate system code developed by CEA, EDF, FRAMATOME-ANP and IRSN for PWR safety analysis, accident management, definition of plant operating procedure and for research and development. It is also used to quantify conservative margins and for licensing. In the framework of Pressurized Water Reactor (PWR) safety studies, Large Break Loss-Of-Coolant Accident (LB LOCA) prediction is still one of the most important and one of the most difficult problem. The three main phases of a LB LOCA are respectively the blowdown, the refilling and the reflooding phases. During the blowdown, the lower plenum voiding results in water entrainment towards the break by steam flowing from the core. Because of the core radial profile, critical heat flux occur but a nonuniform quenching may take place, which results in a 3-D repartition of the energy stored in the core at the beginning of the reflooding. The refilling phase which starts at the accumulator discharge encounters very complex thermalhydraulic phenomena: very strong condensation which induces instabilities, presence of nitrogen degassing from accumulator water which may have an important effect on the transient, countercurrent flow limitation which occurs in the complex geometry of the annular downcomer. The reflooding phase initial conditions in the core are therefore very non-uniform. The presence of buoyancy driven transverse flows below the quench front assures a very efficient mixing between the fuel assemblies. The quench front progression in the hot assemblies is accelerated by pre-cooling due to water cross-flows just above the quench front. Therefore the clad temperature excursion is moderated in the hot assemblies by an increased water carry-over coming partially from colder assemblies. All these multi-dimensional aspects create a very challenging problem for the CATHARE 3-D module. A good prediction of the lower plenum voiding altogether with the amount of Emergency Core Cooling (ECC) water bypassed to the break and of all the cross-flows between the fuel rod assemblies is of prime importance for the whole transient and particularly for the peak clad temperature prediction. Therefore this paper illustrates the CATHARE 3-D module assessment in this context. The separate effect tests and the system effect tests tend to cover all the main phenomena occurring during a LB LOCA and the following assessment results are presented and discussed: - blowdown phase, lower plenum voiding: Prediction of the lower plenum water level assessed against PIERO experiment. - refilling phase: The counter-current flow occurring in the annular downcomer and the amount of ECC water bypassed to the break prediction are assessed against UPTF downcomer refill tests. - reflooding phase: The prediction of the 3-D effects (presence of cross-flows) observed in the core is assessed with the PERICLES 2D experiment. At last, in order to verify the consistency and adequacy of the CATHARE code, the LOFT L2-5 experiment calculation is presented and analysed. (authors)
Availability note (English)
Available in abstract form only, full text entered in this recordAdditional details
Publishing Information
- Imprint Pagination
- 1 p.
- Report number
- INIS-FR--4472
Conference
- Title
- 11. international topical meeting on nuclear reactor thermal hydraulics (Nureth 11)
- Dates
- 2-6 Oct 2005
- Place
- Avignon (France)
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 37050288
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S42: ENGINEERING;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- BLOWDOWN; BYPASSES; C CODES; COMPUTERIZED SIMULATION; CROSSFLOW SYSTEMS; ECCS; FLOW MODELS; LOSS OF COOLANT; PWR TYPE REACTORS; QUENCHING; THERMAL HYDRAULICS
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; ENGINEERED SAFETY SYSTEMS; ENRICHED URANIUM REACTORS; FLUID MECHANICS; HYDRAULICS; MATHEMATICAL MODELS; MECHANICS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR PROTECTION SYSTEMS; REACTORS; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS