%0 Journal Article
%T RELAP5 Analysis of OECD/NEA ROSA Project Experiment Simulating a PWR Loss-of-Feedwater Transient with High-Power Natural Circulation
%A Takeshi Takeda
%A Hideaki Asaka
%A Hideo Nakamura
%J Science and Technology of Nuclear Installations
%D 2012
%I Hindawi Publishing Corporation
%R 10.1155/2012/957285
%X A ROSA/LSTF experiment was conducted for OECD/NEA ROSA Project simulating a PWR loss-of-feedwater (LOFW) transient with specific assumptions of failure of scram that may cause natural circulation with high core power and total failure of high pressure injection system. Auxiliary feedwater (AFW) was provided to well observe the long-term high-power natural circulation. The core power curve was obtained from a RELAP5 code analysis of PWR LOFW transient without scram. The primary and steam generator (SG) secondary-side pressures were maintained, respectively, at around 16 and 8£żMPa by cycle opening of pressurizer (PZR) power-operated relief valve and SG relief valves for a long time. Large-amplitude level oscillation occurred in SG U-tubes for a long time in a form of slow fill and dump while the two-phase natural circulation flow rate gradually decreased with some oscillation. RELAP5 post-test analyses were performed to well understand the observed phenomena by employing a fine-mesh multiple parallel flow channel representation of SG U-tubes with a Wallis counter-current flow limiting correlation at the inlet of U-tubes. The code, however, has remaining problems in proper predictions of the oscillative primary loop flow rate and SG U-tube liquid level as well as PZR liquid level. 1. Introduction High reliability of control rods results in relatively low risk for anticipated transient without scram (ATWS) of pressurized water reactor (PWR). Failure of scram during loss-of-feedwater (LOFW) transient, however, should lead to relatively high core power for a long time and significant thermal-hydraulic responses which may cause degradation in core cooling with gradual loss of primary coolant inventory. Such phenomena include high-power natural circulation with liquid entrainment in hot leg at the inlet of pressurizer (PZR) surge-line, and counter-current flow limiting (CCFL) at the PZR bottom that may hold a large amount of coolant in the PZR, as shown in Figure 1. In the transient following LOFW, power-operated relief valve (PORV) of PZR may continue cycle opening, resulting in loss of primary coolant inventory. The core cooling conditions would then be degraded especially after the natural circulation mode turns into reflux cooling. Figure 1: Thermal-hydraulic phenomena during LOFW transient without scram. A LOFW-induced ATWS experiment was conducted in the LOFT (Loss of Fluid Test) program in the USA and revealed that the primary pressure is kept below about 17.2£żMPa by cycle opening of the PZR PORV and safety valve while the primary fluid temperature
%U http://www.hindawi.com/journals/stni/2012/957285/