Simulation of 1% hot leg SBLOCA with TRACE5

Conference · Sun Jul 01 00:00:00 EDT 2012

OSTI ID:22107794

Gallardo, S.; Querol, A.; Veru, G. ^[1]

---

1. Universitat Poltecnica de Valencia, Cami de Vera s/n, Valencia (Spain)

During a Small Break Loss-of-Coolant Accident (SBLOCA) transient, depressurization can be slow enough to delay the Accumulators (ACC) entry for a long time. Actuation of High Pressure Injection (HPI) system is then necessary in order to maintain the core temperature low enough to avoid core boil off, and consequently avoiding the core level to fall below fuel rods level, thus producing a temperature trip in the fuel cladding. In this frame, the OECD/NEA ROSA Project Test 1.2 (SB-HL-17 in Japan Atomic Energy Agency (JAEA)) has been simulated using the thermal-hydraulic code TRACES. Test 1.2 was performed in the Large Scale Test Facility (LSTF) reproducing a 1% hot leg SBLOCA in a Pressurized Water Reactor (PWR). A comparison between experimental and the main simulated variables was performed to study the effect of important parameters (liquid stratification, geometry and size) to model the break. Finally, in an Emergency Core Cooling System (ECCS) failure scenario, loss of coolant is large enough to produce core boil-off and a Peak Cladding Temperature (PCT) excursion. With this purpose a sensitivity analysis varying the HPI mass flow rate has been performed covering the range between HPI actuation and failure. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22107794
• Resource Relation: Conference: ICAPP '12: 2012 International Congress on Advances in Nuclear Power Plants, Chicago, IL (United States), 24-28 Jun 2012; Other Information: Country of input: France; 8 refs.; Related Information: In: Proceedings of the 2012 International Congress on Advances in Nuclear Power Plants - ICAPP '12| 2799 p.
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
97 MATHEMATICAL METHODS AND COMPUTING
CLADDING
COMPARATIVE EVALUATIONS
COMPUTERIZED SIMULATION
COOLANTS
DEPRESSURIZATION
ECCS
EXCURSIONS
FLOW RATE
FUEL RODS
GEOMETRY
INJECTION
JAEA
LOSS OF COOLANT
NUCLEAR ENERGY
NUCLEAR POWER PLANTS
PRESSURE RANGE MEGA PA 10-100
PWR TYPE REACTORS
SENSITIVITY ANALYSIS
TEST FACILITIES
THERMAL HYDRAULICS