Simulation of a SBLOCA in a hot leg. Scaling considerations and application to a nuclear power plant

Querol, A.; Gallardo, S.; Verdú, G.

doi:10.1016/j.nucengdes.2014.10.006

Cited by 12 publications

(4 citation statements)

References 10 publications

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“…Montero-Mayorga investigated SBLOCA with different break sizes with the best estimate TRACE V5.0 codes, and the main results show the similar trend when compared to Westinghouse NOTRUMP code [21]. Other scholars also conducted similar studies on SBLOCA of loop-type PWR, and some even investigated severe accidents phenomena caused by SBLOCA (the PCCS did not put into use on time) [22][23][24][25][26][27]. However, in this paper, a revised PSIS for GBS Qinshan I has been proposed.…”

Section: Introductionmentioning

confidence: 73%

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

Sun

Wang

et al. 2020

Science and Technology of Nuclear Installations

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Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

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Section: Introductionmentioning

confidence: 73%

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

Sun

Wang

et al. 2020

Science and Technology of Nuclear Installations

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“…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.…”

Section: Lstf Modelmentioning

confidence: 99%

Break location influence in pressure vessel SBLOCA scenarios

Lorduy¹,

Querol²,

Gallardo³

et al. 2021

Braz. J. Rad. Sci.

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The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused on simulating different hypothetical accidental scenarios in the PV using the thermal-hydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios.

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“…5(d) and 7(e)). The differences found in the evolution of the main variables for both installations can be related with the different scaling and/or with the difference in the technology (Querol et al, 2014;Martinez-Quiroga et al, 2014;Toth et al, 2010). Focusing on this latter aspect, the reactor pressure vessel design can greatly influence the evolution of the transient, and some differences are encountered between ROSA/LSTF and PKL facilities.…”

Section: Rosa/lstf and Pkl Comparisonmentioning

confidence: 99%

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

Alberola

Querol

Gallardo

et al. 2016

Nuclear Engineering and Design

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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.

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Simulation of a SBLOCA in a hot leg. Scaling considerations and application to a nuclear power plant

Cited by 12 publications

References 10 publications

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

Break location influence in pressure vessel SBLOCA scenarios

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

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