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Amico P.,Energy Research Inc. | Lubarsky A.,Energy Research Inc. | Lubarsky A.,International Atomic Energy Agency | Kouzmina I.,Energy Research Inc. | And 8 more authors.
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011 | Year: 2011

In accordance with German nuclear regulations, a seismic PSA (SPSA) was performed on Kernkraftwerk Neckarwestheim Unit 2 (GKN II), a PWR located in Germany near Stuttgart. The study was conducted using techniques that comply with both German PSA guidelines and the ANS (now ASME/ANS) standard requirements for SPSA. The study found that the seismic design of the plant is quite high given the seismic hazard at the site. As a result, seismic core damage frequency contributes approximately 1% to total core damage risk of the plant. The risk is dominated by seismically-induced plant shutdown (no loss of offsite power) followed by random failures and human errors, and the dominant seismic events are at the low end of the hazard curve. The results are essentially insensitive to most seismic-related inputs, but are sensitive to the human error probabilities used. The walkdown did indentify few housekeeping items that could compromise the seismic performance of a few components, which the plant is addressing.

Amico P.,Energy Research Inc. | Strohm A.,EnBW | Rattke J.,EnBW
International Topical Meeting on Probabilistic Safety Assessment and Analysis 2011, PSA 2011 | Year: 2011

This paper suggests an approach to seismic HRA that addresses some of the deficiencies of the "shock model" approach commonly used for seismic HRA. The problem with the shock model approach is that it places too much emphasis on the acceleration associated with the seismic event and not enough on the extent of damage caused by the event. Logic suggests that the effects of the acceleration are short-lived as regards human performance (i.e., due to disorientation) and that after a short initial period performance would return essentially to normal other than for the need to deal with the impact of the actual seismic failures. Because of this, the shock model does not adequately allow credit for increased seismic design capacity or long coping times before operator action is required. In this paper, the authors suggest the use of a more context based approach that does account for these influences. The emphasis of this approach is on the overall context under which an action is performed, of which the acceleration is only one part. This allows for better consideration of the broader range of performance influencing factors that result from the actual seismic damage to the plant. The paper presents the methodology and the process for application, and also presents a specific application from the SPSA of the German NPP Kernkraftwerk Neckarwestheim Unit 2 (GKN II). It is concluded that the approach was successful in that application to provide a more realistic treatment of human reliability and so a more accurate risk profile. As such, the approach clearly has promise, but further development is required beyond this first application.

Strohm A.,EnBW | Ehlkes L.,EnBW | Schwarz W.,EnBW | Khatib-Rahbar M.,Energy Research Inc. | And 2 more authors.
10th International Conference on Probabilistic Safety Assessment and Management 2010, PSAM 2010 | Year: 2010

A level-2 Probabilistic Safety Assessment (PSA) is an integrated approach to investigate the progression of severe accidents up to containment failure and release of radionuclides into the environment. The results of a standard level-2 PSA include the frequencies associated with various containment failure modes (release categories) along with the environmental release quantities for various radioisotopes (source terms). The extended level-2 PSA approach discussed in this paper merges the standard level-2 PSA results into an integral metric for risk assessment by estimating the integral risk of activity of radiological release to the immediate vicinity of the plant. Risk is defined as a product of the released activity and the release-category frequency, integrated over all possible release categories. This approach was recently used to assess the risk of severe accidents for the Neckarwestheim Unit 1 (3-loops, 840 MWe) and the Neckarwestheim Unit 2 (4-loops, 1400 MWe) nuclear power plants, which entered commercial operation in 1976 and 1989, respectively. The results have demonstrated that neither the core damage frequency nor the core damage profile necessarily is an adequate indicator of plant risk. Furthermore, neither the absolute frequencies of release categories nor the relative proportions of the release category frequencies necessarily provide a balanced picture of severe accident risk as represented by the integral activity of release.

Madni I.K.,Office of Nuclear Regulatory Research | Khatib-Rahbar M.,Energy Research Inc.
ASME 2011 Small Modular Reactors Symposium, SMR 2011 | Year: 2011

This paper focuses on modeling and phenomenological issues relevant to analysis of severe accidents in integral Pressurized Water Reactors (iPWRs). It identifies relevant thermal-hydraulics, melt progression and fission product release and transport phenomena, and discusses the applicability of the MELCOR computer code to modeling of severe accidents in iPWRs. Areas where the current MELCOR severe accident modeling framework has limitations in the representation of phenomenological processes are identified and examples of possible modeling remedies are discussed. The paper identifies modeling and phenomenological issues that contribute to differences in the calculated reactor coolant system and containment response for iPWRs as compared to traditional PWRs under severe accident conditions. Copyright © 2011 by ASME.

Vasavada S.,Purdue University | Vasavada S.,Energy Research Inc. | Sun X.,Ohio State University | Ishii M.,Purdue University | Duval W.,NASA
Experiments in Fluids | Year: 2011

The results from an experimental study of reduced-gravity two-phase flows are reported in this paper. The experiments were conducted in simulated reducedgravity conditions in a ground-based test facility with a circular test section of 25 mm inner diameter. The flow conditions for which data were acquired lie in the dispersed droplet to slug flow transition and slug flow regime. Local data were acquired for 17 different flow conditions at three axial locations. The acquired data complement and extend those discussed in an earlier paper by the authors (Vasavada et al. in, Exp Fluids 43: 53-75, 2007). The radial profiles and axial changes in the local data are analyzed and discussed in this paper. The area-averaged data, in conjunction with the local data, are discussed to highlight important interaction mechanisms occurring between fluid particles, i.e., drops. The data clearly show the effect of progressive coalescence leading to formation of slug drops. Furthermore, the shape of slug drops in reduced-gravity conditions was observed to be different from that in normal-gravity case. The analyses presented here show the presence of drop coalescence mechanisms that lead to the formation of slug drops and transition from dispersed droplet flow to the slug flow regime. The most likely causes of the coalescence mechanism are random collision of drops driven by turbulence eddies in the continuous phase and wake entrainment of smaller drops that follow preceding larger drops in the wake region. Data from flow conditions in which the breakup mechanism due to impact of turbulent eddies on drops illustrate the disintegration mechanism. © Springer-Verlag 2010.

Helton D.M.,U.S. Nuclear Regulatory Commission | Zavisca M.,Energy Research Incorporated | Khatib-Rahbar M.,Energy Research Incorporated
Safety and Reliability: Methodology and Applications - Proceedings of the European Safety and Reliability Conference, ESREL 2014 | Year: 2015

The US Nuclear Regulatory Commission is developing a site Level 3 Probabilistic Safety Analysis (PSA) for the two operating nuclear power plants, and associated spent nuclear fuel storage facilities at the Vogtle site. The development of this integrated PSA, along with its intended end-uses, shifts the focus in the Level 2 (accident progression and radiological source term) portion of the analysis toward greater realism. This paper describes ways in which this focus affects the analytical approach. Presently in the US, Level 2 PSAs for operating reactors are most frequently performed in either a simplified manner (to support environmental impact studies), or for comparison with a (safety case-related) figure-of-merit. These applications often rely on simplifying assumptions. The present Vogtle analysis is novel in its scope, encompassing all major radiological sources on the site in concert, while assessing a range of hazards and plant operating configurations. As risk assessment technology advances, as computational capabilities increase, and in light of the multi-unit event at Fukushima Daiichi, this type of PSA activity is expected to proliferate. Interactions between the nuclear safety and risk communities at this stage will facilitate the development of more efficient methods. This paper will describe the approach being taken in several areas of the reactor and the spent fuel pool Level 2 PSA to promote realism in the treatment of the overall site risk. These areas include structural failure characterization, modeling of severe accident phenomena, survivability of equipment and instrumentation, and modeling of accident management actions. Whereas a systems-level view of the subject site design and operation would suggest very little dependency between the two reactors and the SFPs, this paper will make the case that after one or more radiological sources have proceeded to a severe accident (i.e., fuel melting) the detailed overall site accident progression becomes very inter-dependent. © 2015 Taylor & Francis Group.

Sawant P.,Energy Research Inc. | Khatib-Rahbar M.,Energy Research Inc. | Moody F.,2125 N. Olive Ave. | Drozd A.,U.S. Nuclear Regulatory Commission
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2011

This paper focuses on the assessment of Advanced Boiling Water Reactor (ABWR) containment pressure-temperature and suppression pool hydrodynamics under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a phenomena identification and ranking table (PIRT) applicable to the ABWR containment response behavior, modeling of pressure-temperature loads using the MELCOR computer code, and analysis of suppression pool hydrodynamics parameters based on a mechanistic one-dimensional hydrodynamics model. A MELCOR 1.8.6 model with detailed nodalization of the ABWR containment is used to perform the containment pressure-temperature calculations following a design basis accident. The best estimate and several sensitivity calculations are performed for the ABWR containment using the suppression pool swell model. The sensitivity calculations demonstrate the influence of key model parameters and assumptions on the suppression pool hydrodynamics response. The comparison of containment pressure-temperature and the suppression pool swell analyses results to those reported in the ABWR licensing calculations showed reasonable agreement. Copyright © 2011 by ASME.

Sawant P.,Energy Research Inc. | Khatib-Rahbar M.,Energy Research Inc.
Nuclear Engineering and Design | Year: 2011

This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements. © 2011 Published by Elsevier B.V.

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