ML19011A436

Lecture 7-1 Retrospective PRA 2019-01-22

ML19011A436
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Issue date:	01/16/2019
From:	Office of Nuclear Regulatory Research
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Nathan Siu 415-0744
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Retrospective PRA Lecture 7-1 1

Course Overview Schedule Wednesday 1/16 Thursday 1/17 Friday 1/18 Tuesday 1/22 Wednesday 1/23 3: Characterizing 7: Learning from Module 1: Introduction Uncertainty 5: Basic Events Operational Events 9: The PRA Frontier L3-1: Probabilistic L5-1: Evidence and L9-1: Challenges for NPP 9:00-9:45 L1-1: What is RIDM?

modeling for NPP PRA estimation L7-1: Retrospective PRA PRA 9:45-10:00 Break Break Break Break Break L1-2: RIDM in the nuclear L3-2: Uncertainty and L5-2: Human Reliability L7-2: Notable events and L9-2: Improved PRA using 10:00-11:00 industry uncertainties Analysis (HRA) lessons for PRA existing technology L9-3: The frontier: grand W1: Risk-informed W2: Characterizing W6: Retrospective 11:00-12:00 thinking uncertainties W4: Bayesian estimation Analysis challenges and advanced methods 12:00-1:30 Lunch Lunch Lunch Lunch Lunch 4: Accident 6: Special Technical 8: Applications and Module 2: PRA Overview 10: Recap Sequence Modeling Topics Challenges L8-1: Risk-informed L2-1: NPP PRA and RIDM: regulatory applications 1:30-2:15 early history L4-1: Initiating events L6-1: Dependent failures L8-2: PRA and RIDM L10-1: Summary and closing remarks infrastructure 2:15-2:30 Break Break Break Break L2-2: NPP PRA models L4-2: Modeling plant and L6-2: Spatial hazards and L8-3: Risk-informed fire Discussion: course 2:30-3:30 and results system response dependencies protection feedback L6-3: Other operational L2-3: PRA and RIDM: W3: Plant systems modes 3:30-4:30 point-counterpoint modeling L6-4: Level 2/3 PRA:

L8-4: Risk communication Open Discussion beyond core damage 4:30-4:45 Break Break Break Break W3: Plant systems W5: External Hazards 4:45-5:30 modeling (cont.) modeling Open Discussion Open Discussion 5:30-6:00 Open Discussion Open Discussion 2

Overview 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

Concept and Use 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?

What can go wrong? What could have gone wrong?

What are the consequences? What would have been the How likely is it? consequences?

How likely was it?

6

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 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

- CCDP = 1 Group, January 1980.

• Based on then-current models. Current estimate ~1E-3 (still a significant precursor). 8

Concept and Use Reminder - NRC Regulatory Functions 9

ASP Program Accident Sequence Precursor Program

• Program recommended by WASH-1400 review group (1978) significant precursor

• Provides risk-informed view of nuclear plant operating experience

- CCDP (events)

- DCDP (conditions) precursor

• Supports reports to Congress*

• Supported by plant-specific Standardized Plant Analysis Risk Licensee Event Reports 1969-2017 models (No significant precursors since 2002)

I. Gifford, C. Hunter, and A. Gilbertson, U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program: 2017 Annual Report, May 2018.

(ADAMS ML18130A856)

• Reports: Abnormal Occurrence, Congressional Budget Justification, Performance and Accountability 10

ASP Program Key Metrics Events Conditions Conditional Core Damage Change in core damage Probability (CCDP): frequency (DCDP):*

l

• Calculated for the duration of the condition. 11

ASP Program Knowledge Check f3 f2 f1 CCDP = ? VA P1 VA P3 P2 12

ASP Program Top U.S. Precursors CCDP/ Event Plant Plant Description DCDP Date Type Cable tray fire caused extensive damage and loss of electrical power to safety Browns Ferry 1 0.4 03/22/1975 BWR systems Failure of non-nuclear instrumentation leads to reactor trip and steam generator Rancho Seco 0.3 03/20/1978 PWR dry out.

Reactor trip results in loss of feedwater with subsequent failure of isolation Oyster Creek 0.03 05/02/1979 BWR condenser.

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 Loss of feedwater; scram; operator error fails emergency feedwater; power-Davis-Besse 0.01 06/09/1985 PWR operated relief valve fails open.

Reactor trip with subsequent failure of high-pressure coolant injection pump to Hatch 2 0.01 06/03/1979 BWR start and reactor core isolation cooling unavailable.

Farley 1 Reactor trip with all emergency feedwater pumps ineffective 0.01 03/25/1978 PWR Blown fuse leads to partial loss of feedwater and subsequent reactor trip; reactor Cooper core isolation cooling and high-pressure coolant injection pump fail to reach rated 0.01 08/03/1977 BWR speed Millstone 2 Loss of offsite power with failure of emergency diesel generator load shed signals 0.01 07/20/1976 PWR Loss of offsite power due to ice storm with failure of emergency diesel generator Haddam Neck 0.01 01/19/1974 PWR service water pump to start 13

ASP Program Most Recent Significant Precursors CCDP/ Event Plant Plant Description DCDP Date Type Reactor pressure vessel head leakage of control rod drive mechanism nozzles, Davis-Besse potential unavailability of sump recirculation due to screen plugging, and potential 0.006 02/27/2002 PWR unavailability of boron precipitation control.

Plant-centered loss of offsite power (transformer ground faults) with an emergency Catawba 2 0.002 02/06/1996 PWR diesel generator unavailable due to maintenance Reactor coolant system blowdown (9,200 gallons) to the refueling water storage Wolf Creek 0.003 09/17/1994 PWR tank High-pressure injection unavailable for one refueling cycle because of inoperable Shearon Harris 0.006 04/03/1991 PWR alternate minimum flow valves Turbine load loss with trip; control rod drive auto insert fails; manual reactor trip; Turkey Point 3 0.001 12/27/1986 PWR power-operated relief valve sticks open CVCS system leak (130 gpm) from the component cooling water/CVCS heat Catawba 1 0.003 06/13/1986 PWR exchanger joint (i.e., small-break loss-of-coolant accident)

Loss of feedwater; scram; operator error fails emergency feedwater; power-Davis-Besse 0.01 06/09/1985 PWR operated relief valve fails open Heating, ventilation, and air conditioning (HVAC) water shorts panel; safety relief Hatch 1 valve fails open; high-pressure coolant injection fails; reactor core isolation cooling 0.002 05/15/1985 BWR unavailable Operator error causes scram; reactor core isolation cooling unavailable; residual La Salle 1 0.002 09/21/1984 BWR heat removal unavailable Salem 1 Trip with automatic reactor trip capability failed 0.005 02/25/1983 PWR 14

Related Activities 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 DCDF < 1E-6

• Part of Reactor Oversight Program DLERF < 1E-7

• Determines significance of findings

- Characterize performance deficiency 1E-6 < DCDF < 1E-5

- Use review panel (if required) 1E-7 < DLERF < 1E-6

- Obtain licensee perspective

- Finalize 1E-5 < DCDF < 1E-4

• Differences from ASP 1E-6 < DLERF < 1E-5

- Supports fault finding and response

- Focuses on a single performance DCDF > 1E-4 deficiency (i.e., not necessarily the DLERF > 1E-5 combined effect of anomalies)

- Results are broad categories (colors) CDF = Core damage frequency LERF = Large early release frequency 16

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 Adapted from U.S. Nuclear Regulatory Commission, NRC Incident Investigation Program, Management Directive 8.3, June 25, 2014. (ADAMS ML13175A294)

- Determines level of reactive inspection 17

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).

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Related Activities IAEA/NEA International Nuclear and Radiological Event Scale (INES)

Level Description

• Tool for communicating event safety significance to the public 7 Major Accident

• Logarithmic scale for severity 6 Serious Accident

• Considers impacts on Accident with Wider People and the environment 5 Consequences Radiological barriers and control Accident with Local Defence in depth 4 Consequences

• Voluntary use by Member States 3 Serious Incident

• Not a notification or reporting system for 2 Incident emergency response 1 Anomaly 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 19

Other Uses Other (Potential) Uses of Retrospective PRA

• Fleet health index

• Alternative method to estimate average CDF 20

Other Uses Integrated ASP Index (IAI)

• Concept

- Use numerical results of ASP analyses to indicate fleet performance

- Increases with number of precursors

- Increases with severity of precursors

• Definition TCY = total calendar years 1 MI = # initiating event precursors

= + MC = # degraded condition precursors CCDP = conditional core damage probability

=1 =1 DCDP = change in core damage probability 21

Other Uses 8.0E-05 7.0E-05 Significant Precursors "Original" Precursors Integrated ASP Index 6.0E-05 LOOP Precursors 5.0E-05 All Other Precursors 4.0E-05 3.0E-05 2.0E-05 1.0E-05 0.0E+00 Calendar Year Adapted from: I. Gifford, C. Hunter, and J. Nakoski, U.S. Nuclear Regulatory Commission Accident Sequence Precursor Program: 2016 Annual Report, May 2017. (ML17153A366) 22

Other Uses Relationship with Fleet CDF?

=

=1 A simple estimator, following Apostolakis and Mosleh (1979):

1 1 1

=

0

=1 =1 =1

• Addresses aleatory uncertainty 1 1

= = = 1

• Same mathematical foundation as basic PRA (Barlow and

=1 =1 Proschan, 1965) 23

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

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 Inputs OpE Program Products Domestic OpE: Industry Influencing Agency programs Daily Event Reports

• Plant Status Reports

• Inspection

• Licensee Event Reports

• Licensing

• Part 21 Reports

• INPO Reports OpE Clearinghouse Informing Stakeholders Domestic OpE: NRC Screening Generic Communications

• Inspection Findings

• OpE Briefings Preliminary Notifications

• Communication COMMunications Regional Project Calls Periodic OpE Newsletter Construction Experience Evaluation OpE Notes Studies/Trends Notable OpE Tech Review Group Report International OpE Application Incident Reporting System (IRS) Taking Regulatory Actions International Nuclear Event Scale (INES) Storage Rulemaking

• Bilateral Exchanges Information Request *

• Available on the public NRC Web Page 26