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[en] The AP 1000 PRA thermal hydraulic simulations were performed with MAAP code, which allows simulating sequences with low computational efforts. On the other hand, the use of best estimate codes allows verifying PRA results as well as obtaining a greater knowledge of the phenomenology and such sequences. The initiating event with the greatest contribution to core damage is Direct Vessel Injection LIne Break (DVILB). This paper presents a review of DVILB sequences of AP 1000 with TRACE code for verifying sequences previously analyzed by Westinghouse with MAAP code. The sequences which configure the DVILB event tree during short term have been simulated. The results obtained confirm the ones obtained in AP 1000 PRA. (Author)
[en] Emergency Operating Procedures (EOP) of US PWRs establish that reactor coolant pumps (RCP) should be tripped during a Small-Break Loss of Coolant Accident (SBLOCA) by the operating crew, provided that the subcooling margin has been lost at the core outlet and the High-Pressure Safety Injection (HPSI) is available. On the other hand, it HPSI is unavailable, RCPs must remain in operation. In this work, it is analyzed human actions in SBLOCA sequences with PHSI failure (and therefore without RCP trip), but with a subsequent HPSI recovery (and therefore with a subsequent RCP trip). The analysis was performed with the TRACE code by means of a model with conservative assumptions. The results show that the HPSI recovery and the subsequent RCP trip can lead the plant to damage conditions it recovery occur when the vessel level is lower. Results also show that such damage can be avoided if only 2 out of 3 RCPs are tripped. (Author)
[en] The objective of the study is the analysis of the appropriateness of the firing of the pumps of the primary and the time available for the beginning of the manual depressurization through steam generators.
[en] Application to the safety function of residual heat removal As part of the project Safety Assessment for Reactor of GEN-IV (SARGEN IV) has been implemented the methodology ISAM from the IAEA to the safety assessment of new sodium reactor designs. Within the ISAM, a new tool to facilitate this assessment is the Objective Provision Tree (OPT) which documents the provisions necessary for each of the levels of defense in depth, as well as for each critical function of security. Due to the design innovations that have sodium reactors, the evaluation of safety and licensing of these reactors requires special considerations. In this work we have analyzed the mechanisms of failure of the safety function concerning the evacuation of waste heat, and have been proposed different provisions for each of the first three levels of defense in depth. The main result of this work is reflected in the elaboration of the OPTs, one for each of the first three levels of defense in depth for the safety of evacuation of residual heat function. These trees represent in a schematic way the provisions necessary to comply with the objectives of each level which are respectively: 1) deviations from normal operation, 2) control of abnormal operation and fault detection and 3) incidental control.
[en] The AP1000 PRA thermal hydraulic simulations were performed with MAAP code, which allows simulating sequences with low computational efforts. On the other hand, the use of best estimate codes allows verifying PRA results as well as obtaining a greater knowledge of the phenomenology of such sequences. The initiating event with the greatest contribution to core damage is Direct Vessel Injection Line Break (DVILB). This paper presents a review of DVILB sequences of AP1000 with TRACE code for verifying sequences previously analyzed by Westinghouse with MAAP code. The sequences which configure the DVILB event tree during short term have been simulated. The results obtained confirm the ones obtained in AP1000 PRA.
[en] Highlights: • Results of ISA for considered sequences endorse EOPs guidance in an original way. • ISA allows to obtain accurate available times for accident management actions. • RCP-trip adequacy and available time for beginning depressurization are evaluated. • ISA minimizes the necessity of expert judgment to perform safety assessment. - Abstract: The integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obtaining the Damage Domain (the region of space of parameters where a safety limit is exceeded) as a function of uncertain parameters (break area) and operator actuation times, and provides to the analyst useful information about the impact of these uncertain parameters in safety concerns. In this work two main issues have been analyzed: the effect of reactor coolant pump trip and the available time for beginning of secondary-side depressurization. The main conclusions are that present Emergency Operating Procedures (EOPs) are adequate for managing this kind of sequences and the ISA methodology is able to take into account time delays and parameter uncertainties
[en] Highlights: • Review of RCP trip issue in case of SBLOCA showing adequacy of present EOPs. • Risk assessment of a SBLOCA deterministic safety analysis by means of ISA methodology. • Evaluation of the probability of damage considering uncertainties in operator actuation times. • Application of ISA methodology to probabilistic safety analysis. • Obtaining of RCP trip available time as function of break size. - Abstract: After the Three Mile Island (TMI) accident, the issue of when to trip the Reactor Coolant Pumps (RCPs) in case of a Small Break Loss of Coolant Accident (SBLOCA) became very important. Several analyses were performed during the 1980s leading to the current Emergency Operating Procedures (EOPs). However these analyses have not been reviewed taking into account that several improvements have been performed in the last thirty years with respect to two phase-flow models, thermal–hydraulics codes and safety assessment methodologies. In this sense, this work has two main objectives: First of all, an assessment of the analyses carried out by Pressurizer Water Reactor (PWR) vendors after the TMI-2 accident with a model of Almaraz Nuclear Power Plant (NPP) for TRACE code (V 5.0 patch 1). On the other hand, Integrated Safety Assessment (ISA) methodology is applied to explore this matter. Such methodology has been developed by the Spanish Nuclear Safety Council (CSN) and it is an adequate method to perform analyses in nuclear safety in which the uncertainties in operator actuation time play an important role. The main conclusions obtained from this work are that, the current EOPs are adequate to manage a SBLOCA sequence in a suitable manner and that ISA methodology is a powerful tool that provides accurate information to the analyst in order to verify the robustness of the EOPs and to perform the safety assessment of both, deterministic and probabilistic safety analysis
[en] Highlights: • Assessment of AP1000 behavior in SBLOCA sequences. • Importance of CMTs and PRHR system for core cooling in case of small break sizes. • Well behavior of the plant in case of availability of half of the total ECCS because of DEDVI. - Abstract: The AP1000® is an advanced pressurized water reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The performing of such systems must be evaluated through the performance of experiments and simulations with a variety of thermal–hydraulic codes. This paper presents the results which has been obtained for different SBLOCA break sizes with the best estimate TRACE V5.0 patch 2 thermal–hydraulic code and their comparison with those obtained by Westinghouse with NOTRUMP code. The main results show that TRACE code predicts a similar trend in all sequences with some differences that are expected to be an issue of the more conservative models and hypothesis assumed in the SBLOCA licensing analysis performed with NOTRUMP. Some particular characteristics of this reactor are also shown in this paper such as the importance of core makeup tank (CMT) and passive residual heat removal (PRHR) system for core cooling in case of small break sizes and the behavior of the plant in case of availability of half of the total passive safety injection systems which is the case of the double-ended direct vessel injection line break (DEDVI)
[en] Highlights: • Assessment of AP1000 behavior in LBLOCA sequences. • AP1000 LBLOCA comparison against standard PWR-3L. • TRACE-DAKOTA application to BEPU analysis. - Abstract: The AP1000® is an advanced Pressurized Water Reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The development of best-estimate codes produced the evolution of conservative safety analysis towards the so-called best-estimate plus uncertainty (BEPU) analysis in order to obtain more realistic results and larger safety margins. In this sense, Westinghouse used for AP1000 Large Break Loss of Coolant Accident (LBLOCA) the so-called Automated Statistical Treatment of Uncertainty Method (ASTRUM) which was developed to address this kind of BEPU analysis. This paper presents a verification of the AP1000 LBLOCA BEPU analysis by means of TRACE V5.0 patch 2 thermal–hydraulic code with the support of DAKOTA code for uncertainty calculations. The results obtained show lower values for the maximum PCT than the ones obtained by Westinghouse. In both cases the results show that AP1000 can mitigate effectively the occurrence of a postulate LBLOCA and to meet the 10CFR50.46 PCT acceptance criteria with enough margin
[en] As part of the participation of the Universidad Politecnica de Madrid research group in CAMP project, an analysis of Medium Break Loss of Coolant Accident (MBLOCA) sequences in a PWR Westinghouse has been performed with TRACE thermal-hydraulic code. The objective of the analysis has been to apply the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), to a thermo-hydraulic analysis of sequences with a cold leg MBLOCA. In order to understand the impact of the availability of the accumulators in this kind of sequences, simulations have been performed both, with available and unavailable accumulators. The ISA methodology, by means of dynamic event trees and path analysis, allows obtaining the damage domain for each sequence of the MBLOCA event tree as a function of the break area and the operator actuation time needed to cool down and depressurize the reactor coolant system by means of steam generators. The results show the capability of the ISA methodology to obtain accurate results that take into account time delays and parameter uncertainties. (author)