The experimental results indicate that in case of high pressure injection (HPI) system failure, the more » rapid depressurization of the steam generators is proved to be an effective way in the depressurization of the reactor coolant system and the core cooling. The break sizes were volumetrically scaled down based on one and all three fully-opened PORVs which is equivalent to 0.46% and 1.37% cold leg flow area. The objectives of this study are to verify the effectiveness of emergency operating procedure and emergency core cooling system on reactor safety. (authors) « lessįour scaled small break loss-of-coolant accident (LOCA) tests simulating the pressurizer power-operated relief valve (PORV) stuck-open accidents and the recovery actions in a pressurized water reactor (PWR) were performed at the INER Integral System Test (IIST) facility.
With this purpose a sensitivity analysis varying the HPI mass flow rate has been performed covering the range between HPI actuation and failure. 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. 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. Test 1.2 was performed in the Large Scale more » Test Facility (LSTF) reproducing a 1% hot leg SBLOCA in a Pressurized Water Reactor (PWR).
#Vtrace steam key generator code#
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. 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. « lessĭuring a Small Break Loss-of-Coolant Accident (SBLOCA) transient, depressurization can be slow enough to delay the Accumulators (ACC) entry for a long time. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates.
#Vtrace steam key generator generator#
The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown more » and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution.
TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Good approximation was obtained between TRACE5 results and experimental data. The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5.
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