ML18284A430
| ML18284A430 | |
| Person / Time | |
|---|---|
| Issue date: | 10/11/2018 |
| From: | Daniel Hudson NRC/RES/DRA |
| To: | |
| Dan Hudson 415-2411 | |
| References | |
| Download: ML18284A430 (10) | |
Text
USNRC Perspectives on Multi-Unit Probabilistic Risk Assessment (MUPRA)
Standard Development Dan Hudson, PhD, CAP, CPH Reliability and Risk Engineer U.S. Nuclear Regulatory Commission
USNRC Site Level 3 PRA Project Objectives
- Develop a contemporary Level 3 PRA that:
Reflects technical advances since NUREG-1150 study.
Addresses risk contributors not previously considered.
- Extract new risk insights to:
Enhance regulatory decision making.
Help focus limited resources on issues most directly related to USNRCs mission to protect public health and safety.
- Improve documentation to make PRA information more accessible, retrievable, and understandable.
- Obtain insight into the technical feasibility and cost of developing contemporary Level 3 PRAs.
USNRC Site Level 3 PRA Project Scope PRA Scope Element Scoping Options Operating Reactor Units Radiological Sources Operating Spent Fuel Pools (SFPs)
Independent Spent Fuel Storage Installation (ISFSI) / Dry Cask Storage Facility At-Power (POS 0)
Operating Reactor Units Low-Power and Shutdown (LPSD)
Nominal Radiological Source Operating SFPs Refueling Outage States Operating Configurations Cask Loading Storage ISFSI / Dry Cask Storage Facility Cask Loading Internal Events Internal Hazards Internal Floods Initiating Event (IE) Internal Fires Hazard Groups Seismic Events External Hazards High Winds Other External Hazards Level 1: Nuclear Fuel Damage Fuel Damage Frequencies PRA End States and Level 2: Radiological Release Categories Radiological Release Frequencies Risk Metrics Frequencies of Offsite Level 3: Offsite Radiological Consequences Radiological Consequences
Integrated Site PRA Key Assumptions
- Single-source PRA model risk insights can be used to prioritize selection of minimal cut sets to be combined.
- Independent multi-source accident scenarios can be screened out.
- Dependent multi-source accident scenarios can be constructed by logically combining minimal cut sets from the single-source PRA models and:
Eliminating logically impossible combinations.
Accounting for effects of inter-source dependencies.
Integrated Site PRA Inputs Integrated Site PRA Technical Approach Limited-Scope Pilot Applications
- Purposes*
Evaluate technical feasibility of implementing proposed approach using existing analytical tools.
Identify potential barriers to implementation.
Identify opportunities for improvement.
- Scope Reactor, At-Power, Internal Events, Level 1 PRA Reactor, At-Power, Seismic Events, Level 1 PRA Reactor, At-Power, Internal Fires, Level 1 PRA Reactor, At-Power and LPSD, Internal Events, Level 1 PRA Reactor, At-Power, Internal Events and Floods, Level 2 PRA
- Key finding For scoping options addressed in the pilot applications, available technology with workarounds can be used to efficiently develop a focused Integrated Site PRA model based on risk insights from single-source models.
- NOTE: No attempt was made to comprehensively identify, characterize, and model inter-source dependencies for each pilot application. Since the main purpose was to evaluate the technical feasibility of the focused approach using existing analytical tools, only a limited set of inter-source dependencies was considered.
Key Technical Challenges
- Treatment of inter-source common-cause failure (CCF) events.
- Treatment of inter-source human reliability analysis (HRA) dependencies.
- Treatment of inter-source seismic correlations.
- Risk aggregation.
- Uncertainty analysis.
Questions for JCNRM Consideration
- What are the potential regulatory concerns related to this issue?
- What is the distinction between multi-unit, multi-module, and multi-source risk?
What are the implications for operating versus new reactors?
- What risk-informed applications are anticipated with respect to each of these types of risks?
What gaps exist in current standards?
When would a standard be needed to support these applications?
How will we distinguish between capability categories for supporting requirements?
- What evidence will be used as the basis for developing new requirements?
What is the potential downside of waiting to obtain additional experience and evidence, considering ongoing efforts?
- What are the expected benefits relative to costs for standard development versus alternative courses of action?
Are there higher priority areas where resources would be better utilized?
Acronyms and Abbreviations CCF Common-Cause Failure HRA Human Reliability Analysis IE Initiating Event ISFSI Independent Spent Fuel Storage Installation JCNRM Joint Committee on Nuclear Risk Management LPSD Low-Power and Shutdown MACCS MELCOR Accident Consequence Code System NPP Nuclear Power Plant POS Plant Operating State PRA Probabilistic Risk Assessment SAPHIRE Systems Analysis Programs for Hands-on Integrated Reliability Evaluations SFP Spent Fuel Pool USNRC U.S. Nuclear Regulatory Commission