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Application of the Integrated Safety Assessment methodology to safety margins. Dynamic Event Trees, Damage Domains and Risk Assessment

The Integrated Safety Assessment (ISA) methodology, developed by the Consejo de Seguridad Nuclear, has been applied to an analysis of Zion NPP for sequences with Loss of the Component Cooling Water System (CCWS).

Accident Management Actions In An Upper-Head Small-Break Loss-Of-Coolant Accident With High-Pressure Safety Injection Failed

Abstract Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of thinning of the vessel head wall at the Davis Besse nuclear power plant reactor indicated the possibility of an SBLOCA in the upper head of the reactor vessel as a result of circumferential cracking of a control rod drive mechanism penetration nozzle—which has cast even greater importance on the study of SBLOCAs. Several experimental tests have been performed at the Large Scale Test Facility to simulate the behavior of a PWR during an upper-head SBLOCA. The last of these tests, Organisation for Economic Co-operation and Development Nuclear Energy Agency Rig of Safety Assessment (OECD/NEA ROSA) Test 6.1, was performed in 2005. This test was simulated with the TRACE 5.0 code, and good agreement with the experimental results was obtained. Additionally, a broad analysis of an upper-head SBLOCA with high-pressure safety injection failed in a Westinghouse PWR was performed taking into account different accident management actions and conditions in order to check their suitability. This issue has been analyzed also in the framework of the OECD/NEA ROSA project and the Code Applications and Maintenance Program (CAMP). The main conclusion is that the current emergency operating procedures for Westinghouse reactor design are adequate for these kinds of sequences, and they do not need to be modified.

Accident Management Actions in a Lower Head SBLOCA With HPSI Failed

In the framework of CAMP and OECD/NEA ROSA projects a broad analysis of Lower Head SBLOCA with High Pressure Safety Injection (HPSI) unavailable in a Westinghouse PWR has been performed. The simulations have been performed with TRACE 5 patch1 and the selected methodology has been the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), which allows obtaining the damage domain as a function of the operator actuation times. The objective of this work is to find the available time for accident management actions by means of ISA methodology. The main conclusions are that present emergency operating procedures are adequate for this kind of sequences and there is not necessary to modify them and also that, the ISA methodology is adequate to analyze safety issues that include time delays uncertainties.Copyright

Application of Integrated Safety Assessment Methodology to Feed and Bleed Sequences

As part of the collaboration between Universidad Politecnica de Madrid (UPM), Indizen Technologies and the Spanish Nuclear Safety Council (CSN), the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of Total Loss of Feed Water sequences in a Westinghouse 3-loop PWR. The ISA methodology allows among others obtaining the Damage Domain, i.e., the region of the space of uncertain parametere where certain damage limit is exceeded. Apart of typical uncertain parameters for the physical model, ISA considers operator actuations times (like RCP trip, AFW recovery and begining of Feed & Bleed…) and /or stochastic uncertain phennomena. Given a particular safety limit, several data of every sequence are necessary in order to obtain the damage exceedance frequency of that limit. These data are obtained from the results of the simulations of transients performed with MAAP code. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.Copyright

Application of Integrated Safety Assessment Methodology to SBO Sequences

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage domain (the region of the uncertain parameters space where the damage limit is exceeded) for each sequence of interest as a function of the operator actuations times (recovery of AC). Given a particular safety limit or damage limit, several data of every sequence are necessary in order to obtain the exceedance frequency of that limit. In this application these data are obtained from the results of the simulations performed with MAAP code transients inside each damage domain and the time-density probability distributions of the manual actions. Several damage limits have been taken into account within the analysis: local cladding damage (PCT>1477 K); local fuel melting (T>2499 K); fuel relocation in lower plenum and vessel failure. Therefore, to every one of these damage variables corresponds a different damage domain. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.Copyright

SIMULATION OF LOSS OF RHRS SEQUENCES IN PKL FACILITY WITH RELAP5 AND TRACE CODES

Abstract An analysis of the PKL midloop tests E3.1 and F2.2 run 2 was performed with the TRACE (TRAC/RELAP Advanced Computational Engine) and RELAP5/MOD3.3 codes. Both tests allow study of the phenomenology and different accident management actions after a loss of the residual heat removal system at midloop conditions with the primary side closed. A comparison of the results obtained with both codes and the experimental data shows that in general, the main phenomena are well reproduced. The good results obtained allow one to confirm that the modelization methodology is adequate for this kind of transient. However, there are still a few phenomena that are not well predicted, like pressurizer water level behavior.

Development of a PWR-W and an AP1000® Containment Building 3D Model With a CFD Code for Best-Estimate Thermal-Hydraulic Analysis

The simulation of design basis accidents in a containment building is usually conducted with a lumped parameter model. The codes normally used by Westinghouse Electric Company (WEC) for that license analysis are WGOTHIC or COCO, which are suitable to provide an adequate estimation of the overall peak temperature and pressure of the containment. However, for the detailed study of the thermal-hydraulic behavior in every room and compartment of the containment building, it could be more convenient to model the containment with a more detailed 3D representation of the geometry of the whole building.The main objective of this project is to obtain a standard PWR Westinghouse as well as an AP1000® containment model for a CFD code to analyze the thermal-hydraulic detailed behavior during a design basis accident. In this paper the development and testing of both containment models is presented.Copyright

Analysis of the Damage Domains of SBO Sequences With RCP Passive Thermal Shutdown

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage domain (the region of the uncertain parameters space where the damage limit is exceeded) for each sequence of interest as a function of the operator actuations times. Given a particular safety limit or damage limit, several data of every sequence are necessary in order to obtain the exceedance frequency of that limit. In this application these data are obtained from the results of the simulations performed with MAAP code transients inside each damage domain and the time-density probability distributions of the manual actions. Damage limits that have been taken into account within this analysis are: local cladding damage (PCT>1477 K); local fuel melting (T>2499 K); fuel relocation in lower plenum and vessel failure. Therefore, to every one of these damage variables corresponds a different damage domain. The operation of the new passive thermal shutdown seals developed by several companies since Fukushima accident is considered in the paper. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.Copyright

Quantitative risk reduction by means of recovery strategies

Abstract After the accident at Fukushima Dai-ichi, considerable efforts were put on enhancing the capability of the Nuclear Power Plants to cope with conditions resulting from the loss of plant safety-related systems. The most widespread solution adopted worldwide has been to define and implement new procedures and emergency actuation plans, the so called FLEX strategies. Among these strategies, there are several recovery strategies which involve the use of portable equipment for accomplishing the safety functions of the unavailable systems. In some cases, these strategies have been devised to be performed concurrently to the usual system recovery procedures included in the EOPs of most NPPs. In this regard, the heat sink recovery after the occurrence of a Total Loss of Feedwater (TLFW) in a Westinghouse 3-loop PWR design is a significant example, and it has been chosen in the present study to assess the quantitative risk reduction provided by the usual and FLEX recovery strategies in a Westinghouse 3-loop PWR design. With this aim, the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to TLFW sequences as part of the collaboration between Technical University of Madrid (UPM), NFQ Solutions and CSN.

Dynamic event trees without success criteria for full spectrum LOCA sequences applying the integrated safety assessment (ISA) methodology

Abstract The integrated safety assessment (ISA) methodology, developed by the Spanish nuclear safety council (CSN), has been applied to the analysis of full spectrum loss of coolant accident (FSLOCA) sequences in a 3-loop pressurized water reactor (PWR). The ISA methodology proposal starts from the unfolding of the dynamic event tree (DET), focusing on the uncertainty of a reduced set of sequence parameters. Outcomes from this step allow assessing the sequence delineation of standard probabilistic safety analysis (PSA) results. For some sequences of interest of the outlined DET, the following ISA methodology steps involve the identification of the damage domain (DD). This is the region of main uncertain parameters space where a safety limit is exceeded during a given sequence. This analysis illustrates the application of this concept, based on transient simulations using MAAP. From the information obtained from the DDs, and considering the time-density probability distributions of human actions and stochastic phenomena occurrence, ISA integrates the dynamic reliability equations proposed to obtain each sequence contribution to the damage exceedance frequency (DEF). The study is then extended to include the uncertainty of subsidiary parameters and, finally, a comparison between the ISA methodology application to FSLOCA and the classical PSA methodology is established.