Module 6 Fire Probabilistic Risk Assessment (PRA) Methodology for Nuclear Power Facilities

Transcription

1 Module 6 Fire Probabilistic Risk Assessment (PRA) Methodology for Nuclear Power Facilities Objectives, Readings, Scope and Assignment Schedule Objectives By the end of this module, students should be able to: Understand fire PRA purpose and process. Appreciate the levels of fire hazard analysis incorporated in the fire PRA. Required Reading Hyslop, J. S., and Kassawar, R. P., (2005). Fire PRA Methodology for Nuclear Power Facilities. NUREG/CR 6850 and EPRI Hyslop, J. S., and Canavan, K. (2010). Fire Probabilistic Risk Assessment Methods Enhancements. NUREG/CR 6850, Supplement 1. NFPA (2010). NFPA 805--Performance-Based Standard for Fire Protection of Light Water Reactor Electric Generating Plants. Quincy, MA, National Fire Protection Association. Vesely, W. E., Goldberg, F. F., Roberts, N. H., Haasl, D. F., (1981). Fault Tree Handbook. NUREG Ruggles, A. E., and Icove, D. (2011). Nuclear Power Plant Fire PRA, Tennessee Industries Week (TIW). PPT file provided. Ruggles, A. E., and Icove, D. (2011). Basics of Fault Tree and Event Tree Analysis, Tennessee Industries Week (TIW). PPT file provided. Suggested References Vigil, R. A. and S. P. Nowlen (1995). An Assessment of Fire Vulnerability for Aged Electrical Relays. Sandia National Laboratories. Albuquerque, New Mexico, U.S. Nuclear Regulatory Commission. Tanaka, T. J., S. P. Nowlen, et al. (1996). Circuit Bridging of Components by Smoke. Sandia National. Laboratory. Albuquerque, New Mexico, U.S. Nuclear Regulatory Commission. Module 6 Assignments Simple Fault Tree Construction (5 points; see below) Cable Vulnerability Assessment (5 points total, see last section of this module for details) US NRC Educational Grant NRC Page 6-1

2 1. Preliminary Activity Fire probability as determined by ignition sources and frequencies can lead to fire events that have capability to compromise safety systems and lead to fuel damage. The probability for fuel damage is represented in a core damage frequency (CDR). Ignition sources that can lead to damage of cable trays with important control, indication or power functions are a significant portion of the plant fire induced risk. Fire damage of switchgear, transformers, motors and other electrical components also contributes. Ignition frequencies are well established for many component types, and nuclear power plant fleet wide data are used to form a composite ignition frequency chart. This composite chart is used in fire PRA assessment. An example of the composite ignition frequency data used for Fire PRA from NUREG CR-6850 is provided in Table 6-1. Table 6-1: Example Partial Listing of Ignition Frequencies (NUREG/CR-6850) An ignition event starts a cascade of events that may lead to core damage. The probability that an ignition event may lead to core damage is determined through probabilistic risk assessment (PRA). PRA methods are known to most nuclear professionals. However, a PowerPoint file (Basics of FT and ET Analysis.PPT) is provided that covers the essential features of fault tree and event tree analysis used in PRA should a tutorial be required. The approach to fire PRA suggested in NUREG/CR-6850, and supplement, is the emphasis of this module. The details of the plant PRA are not covered. The way the PRA drives priorities in fire hazard analysis is emphasized. Only ignition locations that pose significant risk to the plant, as represented by CDF, are given complete fire hazard assessments. NRC has continued to escalate use of performance based methods to set priorities for regulation. The goal is to maximize efficacy of regulation in protecting the welfare of the public. US NRC Educational Grant NRC Page 6-2

3 PRA methods are central to performance based regulation. Fire PRA is currently treated as an addition to the existing plant PRA in NUREG/CR One would expect a more integrated approach to emerge as time passes. As a final note, the PRA methods are increasingly important to assessing risk in insurance actuarial functions, and this may also motivate use of PRA in nuclear power applications going forward. Group activity - For the preliminary activity, the class will discuss the ways a fire ignition, progression and fire protection response might influence the power plant, and how erroneous indications in the control room might confuse operators. Author a group report and each student individually submit their copy to M6 PRA Modeling in the Assignments area in BlackBoard as well as to the instructor or teaching assistant by the due date specified in the Course Calendar. 2. The Fire PRA Process In 2011, the generally accepted step-by-step process for modeling fires was made into an engineering guide by the Society of Fire Protection Engineers. The Guideline for Substantiating a Fire Model for a Given Application serves now as the engineering standard of care for a fire hazard assessment. Over this is another set of recommendations, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, NFPA- 805 (2010), and NUREG/CR-6850 (2005) with supplement (2010). The performance based methods use PRA to prioritize the areas of the plant where detailed fire hazard assessments are beneficial, leading to a tiered approach to engineering evaluation of fire hazard. The most critical areas receive full zone model or computational fluid dynamic treatments using CFAST or FDS. Less critical areas are evaluated using spreadsheet tools. Areas of low impact on plant fire risk may be eliminated from the assessment because no critical components reside in the area, or there is a very low probability of fire. This module presents performance based modeling as it applies to nuclear power plants. The main steps in constructing the fire PRA are provided in Figure 6-1a and Figure 6-1b. The process is formalized, with details offered on documentation and work scope execution for some tasks. The sixteen tasks that comprise the process are expected to consume a few man years of effort for each plant. This man power estimate assumes a plant PRA exists, and that cable routings are documented and available in an electronic and searchable format. The process is also leveraged by an existing fire safe shutdown evaluation for the plant. The fire safe shutdown evaluation is not performance based. US NRC Educational Grant NRC Page 6-3

4 Figure 6-1a: Fire PRA Construction Process (Continued in Figure 6-1b). Figure 6-1b: Fire PRA Construction Process (Continuation from Figure 6-1a). US NRC Educational Grant NRC Page 6-4

5 3. Fire Hazard Calculations Motivated by PRA A PowerPoint file (Nuclear Power Plant Fire PRA.PPT) first used in a pilot short course during Tennessee Industries Week (TIW) is provided. This overview of the fire PRA process shows the progression in fire hazard analysis detail as the relative importance of the fire event to CDF increases. Some areas are eliminated early in the process based on ignition frequency data, and the importance of components in the area to the plant response. This level of assessment escalates to more detailed evaluations of fire zone of influence, and severity factors in scoping fire modeling. Task 11 finally demands detailed fire hazard assessment of areas important to the plant risk posed by fire events. Additional consideration of combined influence of earthquakes with fire, and human reliability occur in the later steps of the process. 4. Assignment Fire PRA for Nuclear Power Plants Individual Activity (5 points): 1) Perform a simple fault tree evaluation and show how an ignition frequency would feed into the probability for failure in the tree. 2) Offer a few examples of how unintended actuations due to cable damage could compromise the plant integrity. 3) List fire ignition frequency items that might be expected to increase with plant age. 4) List fire detection and response components that may deteriorate with age, or for which failure rates may increase with age. CFAST calculations from other course components, such as module 4, can be placed in context with the fire PRA process. Cable endurance can be assessed using fire dynamics tools and cable data using plume centerline temperature predictions based on room geometry and fire HRR. This is provided as an example included in the PowerPoint file, Nuclear Power Plant Fire PRA. Submit your Module 6 Assignment to the M6-PRA Modeling assignment space by the due date for instructor review and grading. US NRC Educational Grant NRC Page 6-5