ML19011A436
| ML19011A436 | |
| Person / Time | |
|---|---|
| Issue date: | 01/16/2019 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| Nathan Siu 415-0744 | |
| Shared Package | |
| ML19011A416 | List:
|
| References | |
| Download: ML19011A436 (26) | |
Text
Retrospective PRA Lecture 7-1 1
Schedule 2
Course Overview Wednesday 1/16 Thursday 1/17 Friday 1/18 Tuesday 1/22 Wednesday 1/23 Module 1: Introduction 3: Characterizing Uncertainty 5: Basic Events 7: Learning from Operational Events 9: The PRA Frontier 9:00-9:45 L1-1: What is RIDM?
L3-1: Probabilistic modeling for NPP PRA L5-1: Evidence and estimation L7-1: Retrospective PRA L9-1: Challenges for NPP PRA 9:45-10:00 Break Break Break Break Break 10:00-11:00 L1-2: RIDM in the nuclear industry L3-2: Uncertainty and uncertainties L5-2: Human Reliability Analysis (HRA)
L7-2: Notable events and lessons for PRA L9-2: Improved PRA using existing technology 11:00-12:00 W1: Risk-informed thinking W2: Characterizing uncertainties W4: Bayesian estimation W6: Retrospective Analysis L9-3: The frontier: grand challenges and advanced methods 12:00-1:30 Lunch Lunch Lunch Lunch Lunch Module 2: PRA Overview 4: Accident Sequence Modeling 6: Special Technical Topics 8: Applications and Challenges 10: Recap 1:30-2:15 L2-1: NPP PRA and RIDM:
early history L4-1: Initiating events L6-1: Dependent failures L8-1: Risk-informed regulatory applications L10-1: Summary and closing remarks L8-2: PRA and RIDM infrastructure 2:15-2:30 Break Break Break Break 2:30-3:30 L2-2: NPP PRA models and results L4-2: Modeling plant and system response L6-2: Spatial hazards and dependencies L8-3: Risk-informed fire protection Discussion: course feedback 3:30-4:30 L2-3: PRA and RIDM:
point-counterpoint W3: Plant systems modeling L6-3: Other operational modes L8-4: Risk communication Open Discussion L6-4: Level 2/3 PRA:
beyond core damage 4:30-4:45 Break Break Break Break 4:45-5:30 Open Discussion W3: Plant systems modeling (cont.)
W5: External Hazards modeling Open Discussion 5:30-6:00 Open Discussion Open Discussion
Learning Objectives
- Retrospective PRA - concept and use
- NRC Accident Sequence Precursor (ASP) program and key results
- Related activities
- Other uses 3
Overview
Resources I. Gifford, C. Hunter, and A. Gilbertson, U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program: 2017 Annual Report, May 2018. (ADAMS ML18130A856)
U.S. Nuclear Regulatory Commission, Accident Sequence Precursor (ASP) Program https://www.nrc.gov/about-nrc/regulatory/research/asp.html N. Siu, et al., Accidents, near misses, and probabilistic analysis:
on the use of CCDPs in enterprise risk monitoring and management, Proceedings of ANS International Topical Meeting on Probabilistic Safety Assessment (PSA 2017), Pittsburgh, PA, September 24-28, 2017.
4 Overview
Other References K.A. Coyne, Risk-Informed Regulation at the U.S. Nuclear Regulatory Commission, April 14, 2016. (ADAMS ML16105A427)
U.S. Nuclear Regulatory Commission, Workshop on the Use of PRA Methodology for the Analysis of Reactor Events and Operational Data, NUREG/CP-0124, 1992.
V.M. Bier (ed.), Accident Sequence Precursors and Probabilistic Risk Analysis, University of Wisconsin Press, Madison, WI, 1998.
Nuclear Energy Agency, Proceedings of the Workshop on Precursor Analysis, NEA/CSNI/R(2003)11, 2003.
J.R. Phimister, V.M. Bier, and H.C. Kunreuther, Accident Precursor Analysis and Management: Reducing Technological Risk Through Diligence, Committee on Precursors, National Academy of Engineering, National Academies Press, New York, 2004.
J.W. Minarick and C.A. Kukielka, Precursors to Potential Severe Core Damage Accidents: 1969-1979, a Status Report, NUREG/CR-2497, June 1982.
G. Apostolakis and A. Mosleh, Expert opinion and statistical evidence: an application to reactor core melt frequency, Nuclear Science and Engineering, 70, 135-149, 1979.
U.S. Nuclear Regulatory Commission, ROP References https://www.nrc.gov/reactors/operating/oversight/program-documents.html U.S. Nuclear Regulatory Commission, NRC Incident Investigation Program, Management Directive 8.3, June 25, 2014. (ADAMS ML13175A294) 5 Overview
What is a Retrospective PRA?
- Preceding lectures address prospective PRA analysis
- identifying and prioritizing possibilities to assist forward-looking decision making
- Retrospective PRA analysis applies a PRA modeling framework and what-if thinking to past events: how close did an incident come to becoming an accident?
6 What can go wrong?
What are the consequences?
How likely is it?
What could have gone wrong?
What would have been the consequences?
How likely was it?
Concept and Use
Why Use Retrospective PRA?
- Support risk-informed prioritization of events for attention and further investigation, possible early warning signals
- Support risk-informed, graded responses to inspection findings
- Provide a different (but still risk-oriented) perspective on plant safety 7
Concept and Use
Early Warning Potential
- Davis-Besse (1977) - LER 346/77-016
- Partial loss of feedwater; stuck-open pressurizer PORV; operators failed to recognize stuck-open PORV
- CCDP = 7x10-2 (analysis ~1982)*
- Three Mile Island 2 (1979) - LER 320/79-012
- Total loss of feedwater; stuck-open pressurizer PORV; operators failed to recognize stuck-open PORV; subsequent operator errors led to core damage
- CCDP = 1 8
Adapted from cover page, M. Rogovin and G.T, Frampton, Jr.,
Three Mile Island: A Report to the Commissioners and to the Public, Nuclear Regulatory Commission Special Review Group, January 1980.
- Based on then-current models. Current estimate ~1E-3 (still a significant precursor).
Concept and Use
Reminder - NRC Regulatory Functions 9
Concept and Use
Accident Sequence Precursor Program Program recommended by WASH-1400 review group (1978)
Provides risk-informed view of nuclear plant operating experience CCDP (events)
DCDP (conditions)
Supports reports to Congress*
Supported by plant-specific Standardized Plant Analysis Risk models 10 Licensee Event Reports 1969-2017 (No significant precursors since 2002) significant precursor I. Gifford, C. Hunter, and A. Gilbertson, U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program: 2017 Annual Report, May 2018.
(ADAMS ML18130A856) precursor
- Reports: Abnormal Occurrence, Congressional Budget Justification, Performance and Accountability ASP Program
Key Metrics Events Conditional Core Damage Probability (CCDP):
Conditions Change in core damage frequency (DCDP):*
11 l
- Calculated for the duration of the condition.
ASP Program
Knowledge Check 12 P3 VA P1 P2 VA CCDP = ?
ASP Program f1 f2 f3
Top U.S. Precursors 13 Plant Description CCDP/
DCDP Event Date Plant Type Browns Ferry 1 Cable tray fire caused extensive damage and loss of electrical power to safety systems 0.4 03/22/1975 BWR Rancho Seco Failure of non-nuclear instrumentation leads to reactor trip and steam generator dry out.
0.3 03/20/1978 PWR Oyster Creek Reactor trip results in loss of feedwater with subsequent failure of isolation condenser.
0.03 05/02/1979 BWR Davis-Besse Both emergency feedwater pumps found inoperable during testing 0.03 12/11/1977 PWR Kewaunee Clogged suction strainers for emergency feedwater pumps 0.03 11/05/1975 PWR Turkey Point 3 Failure of three emergency feedwater pumps to start during test 0.03 05/08/1974 PWR Point Beach 1 Clogged suction strainers for emergency feedwater pumps 0.03 04/07/1974 PWR La Crosse Loss of offsite power due to switchyard fire 0.02 03/24/1971 BWR Davis-Besse Loss of feedwater; scram; operator error fails emergency feedwater; power-operated relief valve fails open.
0.01 06/09/1985 PWR Hatch 2 Reactor trip with subsequent failure of high-pressure coolant injection pump to start and reactor core isolation cooling unavailable.
0.01 06/03/1979 BWR Farley 1 Reactor trip with all emergency feedwater pumps ineffective 0.01 03/25/1978 PWR Cooper Blown fuse leads to partial loss of feedwater and subsequent reactor trip; reactor core isolation cooling and high-pressure coolant injection pump fail to reach rated speed 0.01 08/03/1977 BWR Millstone 2 Loss of offsite power with failure of emergency diesel generator load shed signals 0.01 07/20/1976 PWR Haddam Neck Loss of offsite power due to ice storm with failure of emergency diesel generator service water pump to start 0.01 01/19/1974 PWR ASP Program
Most Recent Significant Precursors 14 Plant Description CCDP/
DCDP Event Date Plant Type Davis-Besse Reactor pressure vessel head leakage of control rod drive mechanism nozzles, potential unavailability of sump recirculation due to screen plugging, and potential unavailability of boron precipitation control.
0.006 02/27/2002 PWR Catawba 2 Plant-centered loss of offsite power (transformer ground faults) with an emergency diesel generator unavailable due to maintenance 0.002 02/06/1996 PWR Wolf Creek Reactor coolant system blowdown (9,200 gallons) to the refueling water storage tank 0.003 09/17/1994 PWR Shearon Harris High-pressure injection unavailable for one refueling cycle because of inoperable alternate minimum flow valves 0.006 04/03/1991 PWR Turkey Point 3 Turbine load loss with trip; control rod drive auto insert fails; manual reactor trip; power-operated relief valve sticks open 0.001 12/27/1986 PWR Catawba 1 CVCS system leak (130 gpm) from the component cooling water/CVCS heat exchanger joint (i.e., small-break loss-of-coolant accident) 0.003 06/13/1986 PWR Davis-Besse Loss of feedwater; scram; operator error fails emergency feedwater; power-operated relief valve fails open 0.01 06/09/1985 PWR Hatch 1 Heating, ventilation, and air conditioning (HVAC) water shorts panel; safety relief valve fails open; high-pressure coolant injection fails; reactor core isolation cooling unavailable 0.002 05/15/1985 BWR La Salle 1 Operator error causes scram; reactor core isolation cooling unavailable; residual heat removal unavailable 0.002 09/21/1984 BWR Salem 1 Trip with automatic reactor trip capability failed 0.005 02/25/1983 PWR ASP Program
Other Precursor Activities
- Past Workshops
- Annapolis, MD, 1992 (NUREG/CP-0124)
- Madison, WI, 1995 (Bier, 1998)
- Brussels, Belgium, 2001 (NEA, 2003)
- Washington, DC, 2003 (Phimister, 2004)
- PSA-Based Event Analysis (PSAEA)
- Annual international workshops led by Belgium
- Exchange results and experiences 15 Related Activities
Significance Determination Process Part of Reactor Oversight Program Determines significance of findings
- Characterize performance deficiency
- Use review panel (if required)
- Obtain licensee perspective
- Finalize Differences from ASP
- Supports fault finding and response
- Focuses on a single performance deficiency (i.e., not necessarily the combined effect of anomalies)
- Results are broad categories (colors) 16 DCDF < 1E-6 DLERF < 1E-7 1E-6 < DCDF < 1E-5 1E-7 < DLERF < 1E-6 1E-5 < DCDF < 1E-4 1E-6 < DLERF < 1E-5 DCDF > 1E-4 DLERF > 1E-5 CDF = Core damage frequency LERF = Large early release frequency Related Activities
Incident Investigation Management Directive (MD) 8.3 Determines NRC response to an incident No additional inspection Special Inspection Team (SIT)
Augmented Inspection Team (AIT)
Incident Inspection Team (IIT)
Differences from ASP
- Quick turnaround
- Determines level of reactive inspection 17 Adapted from U.S. Nuclear Regulatory Commission, NRC Incident Investigation Program, Management Directive 8.3, June 25, 2014. (ADAMS ML13175A294)
Related Activities
Example: Robinson Fire (3/28/2010)
Electrical fault causes fire and subsequent reactor trip with losses of main feedwater (MFW) and reactor coolant pump (RCP) seal injection/cooling (LER 261-10-002)
Incident Response (MD 8.3)
CCDP = 4x10-5 => augmented inspection Initial evaluation recommended a special inspection; loss of RCP seal injection/cooling not known at the time Significance Determination Process (SDP)
Two White findings: licensee performance deficiencies involving inadequate training and procedures.
Five Green findings Accident Sequence Precursor (ASP)
CCDP = 4x10-4 Non-recoverable loss of MFW modeled with RCP seal injection diverted away from RCP seals (unknown to operators) and component cooling water (CCW) isolated via return isolation valve (recovered by operators).
18 Related Activities
IAEA/NEA International Nuclear and Radiological Event Scale (INES)
Level Description 7
Major Accident 6
Serious Accident 5
Accident with Wider Consequences 4
Accident with Local Consequences 3
Serious Incident 2
Incident 1
Anomaly 19 International Atomic Energy Agency, INES: The International Nuclear And Radiological Event Scale, Users Manual, 2008 Edition, © IAEA, 2013.
https://www-pub.iaea.org/books/IAEABooks/10508/INES-The-International-Nuclear-and-Radiological-Event-Scale-User-s-Manual-2008-Edition Tool for communicating event safety significance to the public Logarithmic scale for severity Considers impacts on
People and the environment
Radiological barriers and control
Defence in depth Voluntary use by Member States Not a notification or reporting system for emergency response Related Activities
Other (Potential) Uses of Retrospective PRA
- Fleet health index
- Alternative method to estimate average CDF 20 Other Uses
Integrated ASP Index (IAI) 21
- Concept
- Use numerical results of ASP analyses to indicate fleet performance
- Increases with number of precursors
- Increases with severity of precursors
- Definition
= 1
=1
+
=1
TCY = total calendar years MI = # initiating event precursors MC = # degraded condition precursors CCDP = conditional core damage probability DCDP = change in core damage probability Other Uses
22 0.0E+00 1.0E-05 2.0E-05 3.0E-05 4.0E-05 5.0E-05 6.0E-05 7.0E-05 8.0E-05 Significant Precursors "Original" Precursors LOOP Precursors All Other Precursors Calendar Year Integrated ASP Index Adapted from: I. Gifford, C. Hunter, and J. Nakoski, U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program: 2016 Annual Report, May 2017. (ML17153A366)
Other Uses
Relationship with Fleet CDF?
23 A simple estimator, following Apostolakis and Mosleh (1979):
=
=1
= 1
=1
= 1
=1
= 1
= 1
=1
=1
= 1
=1
= 1
0
Addresses aleatory uncertainty Same mathematical foundation as basic PRA (Barlow and Proschan, 1965)
Other Uses
An Alternative to Standard PRA?
Concept: use statistical estimates of CDF with CCDPs serving as data
- Proposed in early days of precursor analysis (1980s)
- Possibly reviving as part of statistical approaches using actual accidents (e.g., TMI-2, Chernobyl, Fukushima)
Some earlier technical challenges have been addressed (e.g.,
more detailed models)
Continuing technical challenges include:
- Model limitations (shared with prospective PRA)
- Specifying the analysis conditions (the givens): failure memory modeling, neglect of hazard variations
- Incorporating full set of knowledge built into PRAs (e.g., risk from scenarios not involved in actual incident) 24 Other Uses
Comments Retrospective PRA is an extremely valuable source of information generally overlooked by the broader PRA community
- PRA-oriented, structured view of actual events
- Prioritization of issues needing attention Current programmatic challenges to retrospective PRA analyses include:
- Resources spent on arguments over modeling and analysis results
- Questions of added value given existing OpE programs (e.g., NRC OpE Clearinghouse)
- Potential for increased polarization (which is the right approach, vs. what we can learn from different approaches) 25
NRC OpE Clearinghouse 26 Screening Communication Evaluation Application Incident Reporting System (IRS)
International Nuclear Event Scale (INES)
Bilateral Exchanges International OpE Daily Event Reports
- Plant Status Reports
- Licensee Event Reports
- Part 21 Reports
- INPO Reports Domestic OpE: Industry Inspection Findings
- Preliminary Notifications
- Regional Project Calls Construction Experience Studies/Trends Domestic OpE: NRC OpE Clearinghouse Generic Communications
- OpE Briefings COMMunications Periodic OpE Newsletter OpE Notes Notable OpE Tech Review Group Report Informing Stakeholders Inspection
- Licensing
- Influencing Agency programs Rulemaking
- Information Request
- Taking Regulatory Actions Storage OpE Program Inputs Products
- Available on the public NRC Web Page