ML20148D088
| ML20148D088 | |
| Person / Time | |
|---|---|
| Issue date: | 05/27/1997 |
| From: | Carpenter C NRC (Affiliation Not Assigned) |
| To: | Strosnider J NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML19317C316 | List: |
| References | |
| NUDOCS 9705300055 | |
| Download: ML20148D088 (71) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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May 27, 1997 MEMORANDUM T0: Jack R. Strosnider, Chief Materials & Chemical Engineering Branch Division of Engineering
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C.E. Carpenter,Jr.,LeadProjectManager[/
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i FROM:
Materials & Chemical Engineering Branch j
Division of Engineering i
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SUBJECT:
APRIL 29 - 30, 1997, MEETING WITH B0ILING WATER REACTORS VESSEL & INTERNALS PROJECT (BWRVIP) REGARDING BWRVIP-06 On Tuesday, April 29, and Wednesday, April 30, 1997, several members of the NRC staff participated in a public meeting with members of the Boiling Water Reactors Vessel & Internals Project (BWRVIP), the Electric Power Research Institute (EPRI), and the General Electric Company (GE) at the GE facility in j
San Jose, California. The meting was held at the request of the NRC staff to resolve NRC staff concerns regarding the review of the EPRI proprietary report TR-105707, "BWR Vessel and Internals Project, Safety Assessment of BWR Reactor Interni s (BWRVIP-06)," dated October 5, 1995. is the meeting participants. is the BWRVIP April 29, 1997, presentation viewgraphs. is the non-proprietary version of the safety assessment discussed during the first day of the meeting, and Attachment 4 is l
the proprietary version of the safety assessment.
During the April 29, 1997, portion of the meeting, the BWRVIP presented a review of the BWRVIP-06 objectives and scope, and discussed the general approach to the evaluation, including the key assumptions that went into the report. The BWRVIP, for the purposes of the evaluation, assumed that each component or system analyzed was fully failed, and then the consequences of these failures, including the impact on plant response, were determined.
Short-and/or long-term actions to mitigate the consequences were identified, with the acceptance criteria being the capability to achieve a safe shutdown condition. The BWRVIP further assumed that some degradation could be acceptable in the short term, based on (a) being able to detect by plant instrumentation during operation, (b) having redundancy in system and/or component function, or (c) inspecting to minimize the possibility of significant undetected cracking.
The BWRVIP-06 report evaluated the safety function of each component, with an acceptance criteria of being able to achieve safe shutdown, not to maintain the original design.
It describes the component or system, including identifying all potential failure locations, evaluates the consequences of a complete failure at each location, and defines any required actions needed to mitigate the effects of a failure.
The BWRVIP-06 report determined that no q
short-term actions are necessary for postulated failures since all BWRs have a a/0) sufficient le: vel of safety, as based on detectability of component failure by
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existing instrumentation and currently required inspections, structural and/or functional redundancy, and the low probability of a challenging event.
Long-1 term actions were the basis for the component prioritization schedule.
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Jack R. Strosnider The BWRVIP also presented a summary (attachment 2) of a Level 1 quantitative safety assessment of BWR internals that was performed by their contractor Science Applications International Corporation (SAIC).
The BWRVIP committed to providing the full safety assessment to the NRC staff for consideration during the review of the BWRVIP-06 document.
Information regarding this assessment was provided to cognizant staff in the Office of Nuclear Regulatory Research (RES) for inclusion in the confirmatory research that is being performed with regards to degradation of BWR internals.
SAIC's methodology utilized a qualitative assessment to identify potential accident scenarios, and then constructed generic mitigating system fault trees for the various BWR types.
This was done by surveying BWR individual plant evaluations (IPEs) performed to date, identifying important component failure modes, and determining the most common and conservative system alignments for modeling.
Generic data (e.g., generic LOCA frequencies, seismic hazard curves and component seismic fragilities, and other information from NUREG-ll50) was then used to quantify the models.
Unavailability of some information required caused a bounding approach to be taken in performing the quantitative assessment.
The bounding assessment initially assumed that each individual component or system evaluated was failed (e.g., probability of failure equals 1.0), then the frequency of all analyzed accident scenarios was determined.
If the j
frequency of core damage was less than 1.0 X 10'6/ year, no further analyses were performed for that component.
However, if the frequency of core damage was greater than 1.0 X 10/ year, then the individual component or system was
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re-evaluated using the crack growth model to calculate component failure probabilities and the Level 1 quantitative safety assessment was reevaluated to determine the core damage frequency. The Monte Carlo crack growth model that the BWRVIP used requires (a) probability (assumed to be 1.0) that a crack exists in a weld, (b) the crack growth rate data, (c) the critical crack size j
resulting in component failure, and (d) a time period for crack growth.
Two constant crack growth rates were assumed - a pre-1985 rate and a post-1985 rate - to correspond with EPRI chemistry guidelines.
The critical crack size was assumed to extend through the weld thickness, and to range between 70% and j
100% of the weld length.
j The BWRVIP concluded that their bounding assessment had results which indicate that (a) failure of any of the components / systems analyzed will not result in significant safety concerns, (b) confirm the conclusion of the qualitative i
safety assessment that no short-term actions are required, (c) calculated core damage frequencies are conservative, and (d) the generic results are applicable for all BWR/2s through BWR/6s.
During the April 30, 1997, portion of the meeting, the NRC staff and the BWRVIP discussed specific questions related to the review of the BWRVIP-06 report; these questions will be addressed in the BWRVIP-06 safety evaluation.
The NRC staff also discussed briefly the reviews of the BWRVIP Reports " Core Spray Internals Inspection and Flaw Evaluation Guidelines (BWRVIP-18)," dated July 26, 1996, and " Internal Core Spray Piping and Sparger Repair Design
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. Jack R. Strosnider 17, 1996, and the BWRVIP's responses to Criteria (BWRVIP-19)," dated September the NRC staff's Requests for Additional Information (RAls) on these two i
Due to recent collateral reviews performed on other components, the NRC staff recommended that the BWRVIP consider in their RAI responses the documents.
possible effects of other plant transients on the need for core spray.
The BWRVIP staff and the NRC staff agreed to continue to meet to discuss emerging technical issues, status of technical reviews and resource Further meetings will be scheduled as needed.
allocation.
Attachments:
1.
Meeting Participants BWRVIP Viewgraphs 2.
Quantitative Safety Assessment (non-proprietary) 3.
Quantitative Safety Assessment (proprietary).
4.
DISTRIBUTION: see next page I
i s
Jack R. Strosnider.
cc:
Robert Keaten, Executive Chairman Steve Leonard, Technical Chairman BWRVIP. Inspection Task BWRVIP Inspection Task GPU Nuclear Niagara Mohawk Power Company ESBl One Upper Pond Road, Bldg F Post Office Box 63 Parsippany, NJ 07054 Lycoming, NY 13093 Carl Terry, Executive Chairman Robin Dyle, Technical Chairman BWRVIP Assessment Task BWRVIP Assessment Task Niagara Mohawk Power Company Southern Nuclear Operating Co.
Post Office Box 63 Post Office Box 1295 Lycoming, NY 13093 40 Inverness Center Parkway Birmingham, AL 35201 George Jones, Executive Chairman John Wilson, Technical Chairman BWRVIP Mitigation Task BWRVIP Mitigation Task Pennsylvania Power & Light Public Service Electric & Gas Co.
A6-1 N51 Two North Ninth Street Post Office Box 236 j
Allentown, PA 18101 Hancocks Bridge, NJ 08038 Bill Campbell, Executive Chairman Bruce McLeod, Technical Chairman j
BWRVIP Repair Task BWRVIP Repair Task i
Carolina Power and Light Conpany Southern Nuclear Operating Co.
411 Fayetteville Street Post Office Box 1295 Raleigh, NC 27602 40 Inverness Center Parkway Birmingham, AL.35201 l
Warren Bilanin, EPRI BWRVIP Manager Vaughn Wagoner, Technical Chairman Electric Power Research Institute BWRVIP Integration Task 3412 Hillview Ave.
Carolina Power & li.ght Company i
Palo Alto, CA 94304 One Hanover Squa"e 8Cl P.O. Box 1551 Raleigh, NC 27612
e NRC/BWRVIP April 29 - 30, 1997, Meeting Participar.ts NAME ORGANIZATION PHONE FAX R. A. Hermann NRC/NRR/DE/EMCB 301-415-2768 301-415-2444 C. E. Carpenter NRC/NRR/DE/EMCB 301-415-2169 301-415-2444 K. A. Kavanagh NRC/NRR/DSSA/SRXB 301-415-3743 301-415-3577
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J. E. Lyons NRC/NRR/DSSA/SRXB 301-415-2895 301-415-3577 Robin Dyle SNC/BWRVIP 205-992-7121 205-992-5793 Bob McCall Peco Energy Co.
610-640-6389 610-640-6582 Robert Scott Com Ed 630-663-7667 630-663-7171 Bob Carter EPRI 704-547-6019 704-547-6035 Ken Wolfe EPRI 415-855-2578 415-855-2774 i
505-842-7903 505-842-7798 Jeff LaChance SAIC Tom Caine GE 408-925-4047 408-925-1687 Ron Horn GE 408-925-3515 408-925-4175 Sam Ranganath GE 408-925-6825 408-925-5269 Don Knecht GE
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Safety Assessment of BWR Reactor
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Internals BWRVIP-06 I
Presented by:
i Robin Dyle - SNC j
l April 29,1997 E
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i BWRVIP 1
l Purpose of Meeting
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l Review and explain the approach used to perform BWRVIP Safety Assessment (BWRVIP-06)
Present results of the PRA confirmatory evaluation j
Present results of individual component evaluations j
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and final prioritization
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Q&A session with NRC staff f
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BWRVIP j
2
Outline Objectives of BWRVIP-06 Approach Scope Evaluation Procedure Results Prioritization BWRVIP Schedule PRA Evaluation Review of Individual Component Evaluations Conclusions BWRVIP 3
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Objectives of BWRVIP-06
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Perform a qualitative safety assessment of BWR reactor internals and attachments (BWRVIP-06) to assure continuing safe operation (assumes loss of integrity in welded and bolted connections) l 4
t Define short-term and long-term actions needed to ensure safe operation l
Develop overall prioritization of components L
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BWRVIP 4
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Philosophy Individual component locations are assumed to be fully failed
- Failures assumed to be at welds and bolted f
connections since these locations are most logical to l
assume failure
- Based on original component design l
Components evaluated individually and consequences of failure determined
- Impact on accident consequences and plant response l
l BWRVIP l
5 l
1 i
Philosophy (Con't) l Evaluate the safety functions of each component.
t These include:
- Maintaining coolable geometry
- Maintaining control rod insertion times
- Maintaining reactivity control
- Assuring core cooling effectiveness
- Assuring instrumentation availability Acceptance criteria is to achieve safe shutdown, not to maintain original design l
i BWRVIP 6
i Philosophy (Con't)
Failure location can be acceptable for achieving safe shutdown based on:
- Detectability - by plant instrumentation during operation
- Redundancy - system and structural j
- Inspection - minimizes possibility of significant undetected cracking Actions identified as either short-term or long-term l
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I BWRVIP j
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Scope i
Reactor internal components (boundary is nozzle safe-q ends) l
- Safety Related - components that must be relied upon to remain functional during and following design basis event to ensure:
The integrity of the reactor coolant pressure boundary The capability to shut down the reactor and maintain it in a safe shutdown condition l
The capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to 10CFR100 guidelines
- Non Safety-Related - components ~not required to achieve i
- Loose Parts - impact on fuel bundle flow blockage, control rod drive operation and chemical reactions with other internals i
BWRVIP I
6
i Scope (Con't)
Does not include:
f
- Reactor vessel shell and other pressure boundary j
components l
- Consumables (fuel, control rods, etc.)
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Safety Related Components Control Rod Guide Tube SLC/ Core Plate DP CRD Housing / Stub Tube Core Spray Piping l
Core Plate Jet Pump Assembly l.
Core Spray Sparger Orificed Fuel Support LPCI Coupling Shroud j
incore Housing / Dry Tube Access Hole Cover Shroud Support Vessel Instrumentation i
Top Guide i
BWRVIP 10 i
t I
Non-Safety Related Components Steam Dryer Shroud Head and Separators I
Feedwater Sparger i
Surveillance Capsule Holder BWRVIP 11
Evaluation Process L
Description of hardware
- Identification of all potential failure locations Safety assessment
- Evaluation of consequences of complete failure at each location Conclusions
- Definition of required short and long term actions Example - Core Spray
- Refer to Pages 38-50 of BWRVIP-06 BWRVIP e
Results No short term actions required
- All BWR product lines have sufficient level of safety based on:
Detectability of component failure by online instrumentation Structural and/or functional redundancy Detectability of component failure by current inspections Low probability of challenging event Long-term actions
- Basis for component prioritization
- Not required for all components BWRVIP.
13
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Considerat. ions for Pr.iorit.izat. ion Safety function Cracking consequences Operator / system response Detectability Routinely inspected?
Inspection / service history j
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i BWRVIP l
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f Component Prioritization Matrix l
Priority-Component l
High Shroud Core Spray Piping and Sparger j
Shroud Suppod Top Guide Core Plate SLC Medium Jet Pump Assembly j
Low CRD Guide Tube CRD Stub Tube
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In-Core Housing Dry Tube instrument Penetrations Vessel ID Brackets LPCI Coupling BWRVIP
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BWRVIP REPORTS /NRC REVIEW STATUS l.
April 22,1997 l
Submittal Initial NRC RAl/ Response SER
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Report Date Review Process Issued l
Assessment:
- 1. shroud I&E, Rev. 0 9/94 complete complete 9/94
- 2. shroud 1&E, Rev.1 4/95 complete complete 6/95 J
- 3. shroud I&E, Rev. 2 10/96 ongoing
- 5. safety assessment 10/95 5/96 12/96
- 6. GL 92-01, Rev.1, Sup.1 11/95 complete complete 10/96
- 7. shroud reinspection 2/96 5/96 10/96
)
- 8. SS crack growth 3/96 12/96 ongoing
- 9. core spray I&E 7/96 1/97 ongoing
- 10. core plate I&E 12/96 3/97 ongoing
- 11. top guide I&E 12/96 3/97 ongoing
- 12. jet pump riser 12/96 ongoing 4
- 13. SLC I&E 3/97
- 14. shroud support I&E 6/97
- 15. jet pump assembly 1&E 6/97
- 16. CRD guide / stub tube I&E 9/97
- 17. In-core / dry tube I&E 9/97
- 19. Instrument Pen. I&E 12/97
- 20. Internals brackets I&E 12/97 l
- 21. Low alloy crack growth 12/97
- 22. Ni-base crack growth 12/97 l
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Submittal Initial NRC RAI/ Response SER
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Report p_ain Review Process Issued Inspection:
- 1. shroud NDE uncertainty 11/94 complete complete 6/95
- 2. RPV/ internals (BWRVIP-03) 11/95 3/97 ongoing
- 3. core spray visual 4/96 3/97 ongoing
- 4. SLC stds.
8/97
- 5. core spray UT 9/97
- 6. shroud support insp. stds.
12/97'
- 7. jet pump assembly stds.
12/97 8 core plate stds.
12/97
- 9. top guide stds.
12/97 Repair:
- 1. shroud repair design criteria 9/94 complete complete 9/94
- 2. shroud repair design 10/95 complete complete 7/96 l
format / content
- 3. core spray repair design 9/96 1/97 ongoing criteria
- 4. CRD roll expansion 11 /96 ongoing
- 5. core spray replacement 3/97 ongoing design criteria
- 6. core spray overlay repair 2/97
- 7. weldability of irradiated 4/97 materials
- 8. underwater flux core 5/97 welding
- 9. SLC repair criteria 8/97
- 10. jet pump repair criteria 12/97
- 11. top guide repair criteria 12/97
- 12. core plate repair criteria 12/97
Se Submittal Initial NRC RAI/ Response
.SER Report Date Review Process Issued Mitigation:
- 1. credit for mitigation 12/97 i
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PRA Evaluation Perform risk-based PRA analysis of consequences of failure of internal components due to SCC to provide additional confidence in conclusions of safety assessment report Scope: Eight components evalueted
- Those components from safety assessment report whose failure in combination with low probability event could result in increased core damage frequency
- CRD, core plate, core spray piping, CS sparger, jet pump, LPCI coupling, access hole covers, top guide I
BWRVIP 17
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Conclusions All BWR product lines possess a sufficient level of safety based on detection of component failure, structural redundancy and low probability Core damage frequencies below levels of concern Analysis supports BWRVIP work prioritization BWRVIP 18
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Quantitative Safety Assessment Of BWR Reactor Internals l
Presented By:
j JeffLaChance j
Science Applications International Corporation April 29,1997 s
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Presentation Outline i
Introduction Overview Of Probabilistic Risk Assessment (PRA)
Application Of PRA To Assessment Of BWR Reactor Internals Example Quantitative Assessment (Core Spray Piping)
Summary Of Component Analyses /Results 2
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Introduction Purpose Of Quantitative Assessment Is To Confirm Results Of Qualitative Assessment PRA Techniques L sed To Evaluate Potential For Core Damage Due To SCC-i Induced Failure Of Reactor Internals Generic Evaluations Were Performed L sing Conservative System Alignments 3
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Components Quantitatively Evaluated Control Rod Guide Tubes, CRD Housings, & Stto Tuaes Core Plate Core Spray Piping Core Spray Spargers Jet Pump Assembly LPCI Couplings Access Hole Covers Top Guide / Grid 4
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Brief History Of PRA l
- First Major Application Was The Reactor Safety Study In 1975 (WASH-1400)
- NUREG-1150 Study Of Five Plants Using Updated Methodologies And Data
- Individual Plant Examinations (IPEs) Required-i By GL 88-20
- NRC Policy Statement On The Use Of PRA In Regulatory Applications
Overview Of PRA A PRA Involves Three Sequei1tial Parts Or l
" Levels"
- Level 1 - Identification And Quantification Of The Sequence Of Events Leading To Core Damage
- Level 2 - Evaluation And Quantification Of The Mechanisms And Amounts Of Radioactive Material Released From Containment
- Level 3 - Evaluation Of The Consequences To The Public 6
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I PRA Process i
Level 1 Analysis Fault Tree 1
Analysis Initiating Event Event Tree Level 2 Level 3 i
Ana ysis Analysis Analysis Analysis Human Reliability Data Analysis Analysis i
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l Initiating Events Initiating Event Identification
- General Transients l
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- Loss-Of Coolant Accidents (LOCAs)
- External Events (e.g., Seismic, Tornadoes, etc.)
Initiating Events Grouped According To:
- Similar Plant Re~sponse
- Same Requirements For Mitigation j
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Accident Sequence Models k
Sequence OF Events Leading To Core Damage Depicted Using Event Trees
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- Binary Model Showing Success Or Failure Of Mitigating Systems And Operator Actions Mitigating System Failure Modes Determined i
Using Fault Trees
- Component Failures
- Human Errors
- External Event Interactions 9
Example Event Tree ACCIDENT REACTOR SRVs OPEN TO COOLANT DECAY HEAT SEO. NO.
SEQUENCE INITIATING SUCCESSFULLY REUEVE SUPPLIED TO REMOVED FROM OUTCOME EVENT OCCURS SCRAMS VESSEL VESSEL CONTAINMENT PRESSURE IMTIATOR RPS SRV INJECTION DHR 1
CORE OK 2
CONTAINMENT FAILURE 3
CORE DAMAGE 4
VESSEL RUPTURE 5
ATWS 10
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l Quantitative Assessment Methodology Utilized Qualitative Assessment To Identify Potential Accident Scenarios Of Concern
- Seismic-Induced Anticipated Transient Without Scram (ATWS)
- Large Break Recirculation Line LOCA Generic Event Trees Constructed For Various Vintages of BWRs
- Typically BWR/2, BWR/3-/4s, & BWR/5-/6s 12
i Quantitative Assessment Methodology (cont.)
Generic Mitigating System Fault Trees
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Constructed
- Survey of BWR IPEs Performed
- Most Common And Conservative System Alignments Chosen For Modeling
- Available BWR IPEs Reviewed To Identify Important Component Failure Modes To Include In Fault Trees-l l
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Quantitative Assessment Methodology (cont.)
Modeling Of Seismic Events Required Special Considerations i
- Seismic Hazard Curve Yielding Highest Frequency Earthquakes Selected For Plant In Each Group (Discretized Into 15 Intervals) i L
- Seismic-Induced Component Failures Added To Fault Trees
- For SCC-Degraded Components, Fragilities Assumed To Be 1.0 For All Levels Of Earthquakes
- Frequency Of Core Damage From Seismic-Induced Failures Quantified For Accelerations Up To 1.0 g 14 l
1-i
Quantitative Assessment Methodology (cont.)
l Generic Data Used To Quantify Models
- Generic LOCA Frequencies
- Seismic Hazard Curves
- Random Component Failure Data And Common-Cause Failure Data (NUREG-1150) l
- Component Seismic Fragilities
- Human Error Probabilities (NUREG-1150) 15
l r
l Modeling Of SCC-Degraded Component Failures Degraded Component Modeling Requires Knowledge Of:
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- How Many Components Must Fail To Result In Adverse Condition
- Depth And Length Of Crack Penetration Required For Individual Component Failure Unavailability OfInformation Resulted In l
Bounding Approach In Performing The i
Quantitative Assessment 16
- i t
i Bounding Assessment Approach l
Quantification Of Models Performed First With Probability Of SCC-Degraded Components Failures Set i
To 1.0 If The Frequencies From All Accident Sequences <1E-6/yr, No Further Analysis Required i
If >1E-6/yr, Crack Growth Model Used To Calculate Component Failure Probabilities & PRA Models Reevaluated Sensitivity Calculations Performed Results Indicate Which Components Are More Important To Safety Based On Prevention Of Core Damage 17
Crack Growth Model Monte Carlo Crack Growth Model Used To Determine Probability Of Component Failure Due To SCC Model Requires:
- Probability Crack Exists In a Weld (Conservatively Assumed To Be 1.0)
- Crack Growth Rate Data j
- Critical Crack Size Resulting In Component Failure l
- Time Period For Crack Growth i
18 I
l Crack Growth Rates Two Crack Growth Rates For Austentic Stainless Steel Used: Before And After 1985 Water Chemistry Changes Pre-1985 Rate Applied For Period Ranging From Time Earliest Plant In Each BWR Vintage Reached Criticality To 1985 Post-1985 Rate Applied For 10 Year Period Rates Assumed To Be Constant Over These Periods 19
Crack Growth Rates (cont.)
Empirical y-Based Correlation Developed For BWRVIP Used To Detennine Rates Assuming Following Conditions:
- Stress Intensity Factor = 30 Ksi* V inch
- Water Conductivity = 0.15 S/cm (post-1985)
= 0.5 pS/cm (pre-1985)
- Electrochemical Potential = 200mV
- Test Temperature = 561 K Sensitivity Studies Perfonned For Different Stress Intensity Factors 20 e
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i Crack Growth Rate Correlation f
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l PROPRIETARY INFORMATION r
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I Calculated Crack Growth Rate Lognormal Distributions l
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l PROPRIETARY INFORMATION 1
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Critical Crack Sizes i
Crack Growth Along The Lerigth Of a Weld i
Assumed Required For Component Failure Cracks Assumed To Extend Through The Weld Thickness 1
Critical Crack Size Information Did Not Exist For Components OfInterest j
Critical Crack Sizes Thus Assumed To j
Range Between 70-100% Of Weld Length j
23 l
i Monte Carlo Method i
Randomly Samples Crack Growth Rate Distributions (Both Pre- & Post-1985)
Applied To Both Tips Of a Crack Resulting Crack Size Categorized And Compared To j
Critical Crack Size Required For Component Failure Process Repeated 1000 Times Number Of Samples Exceeding Critical Crack Size Divided By 1000 To Obtain Probability Of Exceeding Critical Crack Size Probability Used in PRA Models j
i 24 l
.t
h Example Quantitative Assessment - Core Spray Piping j
Concern Is That SCC-Degraded Core Spray l
Piping Will Fail During Large Recirculation Line Break l
- PRA Model Descriptions j
Break Size Assumptions l
Event Tree i
- Evaluation Results 25
L Potential Core Spray Piping Failure Locations' t
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PROPRIETARY INFORMATION i
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...l BWR Line Dependency
)
BWR/2s Do Not Have Jet Pumps And Rely On Core Spray (CS) For Core Cooling BWR/3s And Most BWR/4s Inject LPCIInto
)
Recirculation Lines Which Ensures Long-Term
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Steam Cooling Of Upper Core (CS Not Required)
)
i BWR/5s & /6s (And Some BWR/4s) Require CS To Ensure Long-Term Core Cooling Even If LPCI Is Successful 27 l
Critical Recirculation Line Break Sizes For BWR/2s, Critical Break Size Is Based On Ability To Reflood The Core Given CS Flow Bypassed Into Annulus Area
- ~0.5 ft If Flow From One CS Train Available 2
2
- ~1.0 ft If Flow From Two CS Trains Available i
For BWR/5s, /6s, & Few /4s; Jet Pump Flow Area Limit Ability To Reflood Core With 3 LPCI Trains For Break j
Sizes >1.5 ft2 Frequency Of Breaks Of These Sizes Assigned Value Of 7.5E-6/yr Based On Survey Of Data 28 i
.l L ---
i Event Tree A.ssumptions BWR/2 Event Tree Assumptions
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t
- Credits Altcinate Coolant Injection Systems (Raw l
Water Or Firewater) Based On IPE Models BWR/4 & BWR/5-/6 Assumptions
- No Credit For Alternate Coolant Injection Systems (Conservative For Some Plants)
- Impact Of CS Pipe. Break On Ability To Inject Boron i
During ATWS Not Developed Due To Low Frequency Of LOCA-ATWS Event
- Failure Of LPCI Couplings From SCC Included In Evaluation 29
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BWR/2 Large Recirculation Line l
LOCA Event Tree t
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t PROPRIETARY INFORMATION 1
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BWR/4 Large Recirculation Line LOCA Event Tree t
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i PROPRIETARY INFORMATION i
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1 BWR/5 & /6 Large Recirculation Line LOCA Event Tree PROPRIETARY INFORMATION I
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Core Spray Piping Results i
i PROPRIETARY INFORMATION
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Study Utilized Conservative Assumptions j
PRA Model Assumptions:
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- Conservative System Alignments
- Alternate Mitigating Systems Not Credited l
- LOCA During An ATWS Treated As Unmitigatable
- Full Power ATWS Modeled i
- Most Limiting Seismic Hazard Curves Used
- Seismic-Induced Failure Of Components Modeled
- Seismic Failure Of SCC-Degraded Components Assumed To Occur With Probability Of 1.0 42 i
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i Study Utilized Conservative Assumptions (cont.)
Crack Growth Model Assumptions:
- Crack Growth Evaluated Using Bounding Conditions
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Expected To Encompass All Vessel Components
- Crack Growth Rates Prior To 1985 Are 3.4. Times Higher Than Current Rates
- Crack Growth Evaluated For Oldest Plant In Each l
Group And Applied Generically To All Plants In Group
- Crack Growth Applied Along Length Of Weld And Assumed To Be Completely Through Weld Thickness
- Cracks Allowed To Grow From Both Ends i
43
Component Prioritization Based On Quantitative Assessment High Priority Components:
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- Top Guide In BWR/3s And Those Early BWR/4s
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Without Brackets And Wedges
- Core Spray Piping And Spargers In All BWRs Medium Priority Components:
- Core Plate Components In BWR/2s Through BWR/5s t
- CRDs For All BWRs Low Priority Components
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- Jet Pump Assemblies
- Access Hole Covers j
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f Conclusions Results Indicate That Failure Of Any Of The 8 Reactor Internal Components Analyzed Due To SCC Will Not Result In Significant Safety Concerns j
Results Also Confirm The Conclusion Of Qualitative l
Safety Assessment That No Short-Term Actions Are Required j
Core Damage Frequencies Calculated In This Study Are Conservative Due To Use Of Conservative Models, 1
Assumptions, And Estimates Of SCC-Induced Failure i
Probabilities l
Although Performed At a Generic Level, The Results Are Applicable For All BWR/2s Through BWR/6s 45 j
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