Abstract
The present paper deals with the analytical study of the PKL experiment G3.1 performed using the TRACE code (version 5.0 patch1). The test G3.1 simulates a fast cooldown transient, namely, a main steam line break. This leads to a strong asymmetry caused by an increase of the heat transfer from the primary to the secondary side that induces a fast cooldown transient on the primary side-affected loop. The asymmetric overcooling effect requires an assessment of the reactor pressure vessel integrity considering PTS (pressurized thermal shock) and an assessment of potential recriticality following entrainment of colder water into the core area. The aim of this work is the qualification of the heat transfer capabilities of the TRACE code from primary to secondary side in the intact and affected steam generators (SGs) during the rapid depressurization and the boiloff in the affected SG against experimental data.
Highlights
Experimental programmes in scaled-down integral test facilities are conducted for solving open issues for current nuclear power plants, for demonstrating the technical feasibility of innovative designs, and for generating reference databases in order to support codes development and assessment [1]
The OECD/NEA CSNI PKL-2 project (2008–2012) is aimed at studying selected accident scenario at system level and understanding the thermal-hydraulic phenomena and processes occurring in pressurized water reactor design as well as validating and improving complex thermal-hydraulic system codes used in safety analysis
The OECD/NEA CSNI PKL-2 (2008–2012) is aimed at studying selected accident scenario at system level and understanding the thermal-hydraulic phenomena and processes occurring in pressurized water reactor design as well as validating and improving complex thermal-hydraulic system codes used in safety analysis
Summary
Experimental programmes in scaled-down integral test facilities are conducted for solving open issues for current nuclear power plants, for demonstrating the technical feasibility of innovative designs, and for generating reference databases in order to support codes development and assessment [1]. Experimental data are fundamental for demonstrating the reliability of computer codes in simulating the behaviour of an NPP (nuclear power plant) during a postulated accident scenario: in general, this is a regulatory requirement [2]. The design of the experiment involves two phases: the first is based on the 0.1A break in main steam line as initiating event and the second one consists of the ECCS injections by means of the HPIS connected with the cold legs number 1 and number 4. Another 0.1A main steam line break test was already performed in PKL facility in 1989: the test B5.1
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