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[en] Highlights: • Verification of AP1000 DVILB event tree with TRACE code. • Analysis of AP1000 DVILB event tree by considering European RNS design. • AP1000 CDF reduction due to European RNS design. - Abstract: As part of AP1000® design adaptation to European Utility Requirements, Westinghouse has carried out several modifications of the standard AP1000 design. One of these modifications has been the physical separation of the normal residual heat removal system (RNS) into two independent trains which improves the plant defense in depth. The AP1000 Probabilistic Risk Assessment, assumes in the standard design that in the event of direct vessel injection line break (DVILB) the RNS cannot inject efficiently, and therefore it is necessary to depressurize the Reactor Coolant System until atmospheric conditions in order to allow injection through the IRWST and to achieve safe conditions. This work presents the analysis performed for DVILB event tree taking into account the RNS modification and the effectiveness of RNS injection. With this purpose, different possibilities of DVILB event tree have been simulated with TRACE V5.0 patch 2 thermal–hydraulic code. The results show that some sequences that were classified as damage with the standard RNS design became success with the consideration of RNS improvement and therefore the contribution of DVILB to the total core damage frequency can be minimized.
[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 aim of this work is to test the capabilities of the new tool of uncertainty incorporated into SNAP by simulating experiments with TRACE code and compare these with the results obtained by the same simulations with uncertainty calculation performed with the tool SUSA.
[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] This paper addresses the most important issues regarding design basis threat concept within the field of nuclear security, international recommendations, threat evaluation and the steps needed for it. Finally, the process carried out in Spain for defining the design basis threat and its direct consequences in the regulation is presented.
[es]El artículo aborda las cuestiones más importantes dentro del concepto de definición de amenaza base de diseño en el campo de la seguridad física; recomendaciones internacionales, evaluación de la amenaza y pasos necesarios para una determinación de la misma. Finalmente, se hace referencia al proceso llevado a cabo en España para la definición de la amenaza y cual han sido las consecuencias directas de la implantación de este concepto en la regulación.
[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] Highlights: • Verification of AP1000 low-margin sequences evaluation process. • Analysis of AP1000 low-margin bounding sequences. • Comparison of TRACE results with respect to NOTRUMP and WCOBRA/TRAC. - Abstract: The Westinghouse AP1000® reactor is an advanced design whose safety systems are mainly passive safety systems. Due to the passive nature of the safety related systems and its dependency on small changes on certain variables (e.g. pressure, friction coefficients) together with the use of a simplified code like MAAP in Probabilistic Risk Assessment (PRA) analyses, it makes necessary to confirm that when core cooling is achieved, thermal-hydraulic (T/H) uncertainties are bounded. The T/H uncertainty evaluation process performed by Westinghouse Electric Company (WEC) identified the low-margin sequences (core uncovery) by expanding PRA Event Trees (ETs). The expanded ETs allowed finding the low-margin risk-important sequences and then a set of low-margin bounding sequences was selected. Then detailed safety analysis methodologies were applied in order to evaluate the bounding sequences demonstrating that the T/H uncertainties were bounded. The Universidad Politécnica de Madrid group has verified the low-margin bounding sequences obtained by WEC with the best-estimate TRACE code in order to verify the previous results and also to study the phenomenology of such sequences through a best-estimate code. This paper presents the results obtained for short-term low-margin bounding sequences. In general, TRACE results do not present important discrepancies with respect to NOTRUMP and WCOBRA/TRAC results although TRACE presents lower values for Peak Cladding Temperature. This analysis has allowed to verify the AP1000 thermal-hydraulic bounding evaluation process performed by WEC for the low-margin risk-important sequences with TRACE code.
[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)