Post-test analysis of the ROSA/LSTF and PKL counterpart test

Abstract

h i g h l i g h t s• TRACE modelization for PKL and ROSA/LSTF installations.• Secondary-side depressurization as accident management action.• CET vs PCT relation.• Analysis of differences in the vessel models. a b s t r a c t Experimental facilities are scaled models of commercial nuclear power plants, and are of great importance to improve nuclear power plants safety. Thus, the results obtained in the experiments undertaken in such facilities are essential to develop and improve the models implemented in the thermal-hydraulic codes, which are used in safety analysis. The experiments and inter-comparisons of the simulated results are usually performed in the frame of international programmes in which different groups of several countries simulate the behaviour of the plant under the accidental conditions established, using different codes and models. The results obtained are compared and studied to improve the knowledge on codes performance and nuclear safety.Thus, the Nuclear Energy Agency (NEA), in the nuclear safety work area, auspices several programmes which involve experiments in different experimental facilities. Among the experiments proposed in NEA programmes, one on them consisted of performing a counterpart test between ROSA/LSTF and PKL facilities, with the main objective of determining the effectiveness of late accident management actions in a small break loss of coolant accident (SBLOCA). This study was proposed as a result of the conclusion obtained by the NEA Working Group on the Analysis and Management of Accidents, which analyzed different installations and observed differences in the measurements of core exit temperature (CET) and maximum peak cladding temperature (PCT). In particular, the transient consists of a small break loss of coolant accident (SBLOCA) in a hot leg with additional failure of safety systems but with accident management measures (AM), consisting of a fast secondary-side depressurization, activated by the CET. The paper presents the results obtained in the simulations for both installations using TRACE, observing, in general, a good agreement with the experiments. However, ROSA/LSTF calculations underestimated the maximum PCT value, what might be explained by the higher core level predicted in the simulation compared with the experiment. In PKL calculations, PCT maximum value is slightly higher than in the experiment, and the core level predicted is lower. In the comparison of the evolution of both installations a different timing in the transient events is observed, due to the difference in the pressure vessel design. Thus, when PKL vessel is modified with some of the ROSA/LSTF features, the evolution of the new PKL model behaviour is closer the one observed in ROSA/LSTF calculations.

Uncertainty analysis of PKL SBLOCA G7.1 test simulation using TRACE with Wilks and GAM surrogate methods

“…PCT and CET evolutions are well reproduced by TRACE, although the CET is slightly higher in TRACE calculation than in the experimental data. In general, the plant behavior observed in the simulation agree with the experimental results, as explained in some more detail in reference (Carlos et al, 2016). So, this simulation is taken as base case to undertake the uncertainty analysis in the following sections, considering the PCTmax as output variable and adopting the methodology introduced in section 2.…”

Break location influence in pressure vessel SBLOCA scenarios

“…This fact led up in the past to perform different studies to define the distribution of nodes in the vessel or the U-tube bundles of the steam generators. Previous works [7,8,11,12,13] have served to verify the model by reproducing different ROSA experiments. Specifically, the capability of the model to correctly reproduce Test 6.1 has been confirmed, comparing simulated and experimental results.…”

RELAP5 Simulation of PKL Facility Experiments under Midloop Conditions

“…The PKL tests results have also been used for preparation and verification of procedures described in the operating manuals and for answering questions from the regulatory bodies. In particular, series E, F, and G of PKL III program include research on the inadvertent boron dilution events, the effect of primary coolant inventory in transients under OM conditions, and analysis of asymmetric cooldown transients [6,7,[9][10][11].…”