ML20141C360
| ML20141C360 | |
| Person / Time | |
|---|---|
| Issue date: | 05/12/1997 |
| From: | Ali S NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9705190030 | |
| Download: ML20141C360 (111) | |
Text
{{#Wiki_filter:t ,pnuoug$ [ UNITED STATES - NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20565-0001 %..... [ May 12, 1997 MEMORANDUM T0: The Files FROM: hSyedAli,StructuralEngineer Civil Engineering and Geosciences Branch Division of Enginoering, NRR
SUBJECT:
SUMMARY
OF MEETIN3 WITH EPRI/NEI ON RISK-INFORMED ISI DATE: April 29, 1997 Lk iTION: U.S. Nuclear Regulatory Commission 11545 Rockville Pike, Rockville, MD 20852 Staff members from NRR and RES met with representatives of the Electric Power Research Institute (EPRI), Nuclear Energy Instituie (NEl) and the industry on April 29, 1997, at the NRC headquarters in Rockvilh. Maryland to discuss technical details of the EPRI sponsored Risk-Informed Inservice Inspection (RI-ISI) methodology. A list of attendees is included as Attachment 1, "EPRI Risk-Informed Inservice Inspection" as Attachment 2, " Technical Basis of ~ Approach" as Attachment 3, "PSA Principles for Consequences" as Attachment 4, " Example of EPRI RI-ISI Application" as Attachment 5, Pipe Failure Potential l via Degradation Mechanism Assessment" as Attachment 6, and " Risk Impact - A Qualitative Examination" as Attachment 7. The industry has sinitted two methodologies to the staff for the implementation of RI-ISI of piping. One methodology has been jointly j developed by ASME Research and Westinghouse Owners Group (WOG) and the other methodology is being sponsored by EPRI. Each of the two methodologies have i their respective pilot plants. In addition, ASME is working on three Code Cases for alternate examination requirements to ASME Section XI, Division 1 for piping welds. Code Case N-577 is based on the WOG methodology, Code Case N-578 is based on the EPRI methodology, and the Code Cases N-560 (for ASME Class 1 piping only) is based on the EPRI methodology. There is currently an effort underway to revise Code Case N-560 to include WOG methodology also. EPRI made an initial presentation of its methodology on September 20, 1996. The purpose of this meeting was for EPRI to present more details of its methodology and address the concerns raised in the September 20, 1996 meeting. EPRI presented the technical details of its RI-ISI process including a discussion of consequence and failure potential assessment due to pipe failures, which are then combined to categorize piping segments for ISI. 160071 DEDN ' CONTACT: Syed Ali, DE/ECGB y CdQ h 415-2776 3 pt 9705190030 970512 P M Qn E mycgN7strg5 PDR TOPRP EXIE p-@ pf V 11T9
i The Files l EPRI presented a process to categorize the consequences of piping segments based primarily on a qualitative analysis of plant specific success paths for mitigating accident sequences. EPRI stated that it was possible to perform J the analysis under the quality er,surance guidelines applied to other deterministic analyses used for licensing purposes. A plant specific Probabilistic Risk Assessment (PRA) could be used to provide quantitative support to the consequence categorization, but EPRI envisions PRA quantification to be used generically to calibrate the process, and to resolve complex interactions in specific instances. Industry representatives also provided an executive update of pilot studies for Fitzpatrick and ANO-2 plants. Industry representatives stated that j utility submittals for pilot plants are scheduled for September 1997. NRC staff commented that the performance monitoring feedback needs to be included in the EPRI RI-ISI process. NRC staff also indicated that the RI-ISI process must address risk as discussed in the Draft Regulatory Guide DG-1061. The staff members provided additional comments on the details of the EPRI methodology which will be documented in the Request of Additional Information (RAI) to be transmitted to NEI/EPRI. The meeting concluded with the agreement that additional public meetings will l be held to discuss further details of EPRI methodology and its application to l pilot plants. Attachments: As stated. DISTRIBUTION: File Center (T-5 C3) PDR (LL-6) ECGB RF NRR Mailroom NRC Meeting Participants AThadani GHolahan MHayfield JJohnson GMillman Glainas JMurphy l WHodges LShao JStrosnider KWichman l RJones JAustin RWessman DFischer DISK / DOCUMENT NAME: G:\\ALI\\ SUM 04978 To rsceive a copy of this document, indicate in the box: "C" s 4opy,w/o attachment; "E" = Copy w/ attachment; "N" s No Copy 0FC ECGB:DE 6-ECGB:DE Ebh l NAME SALI.160 P0 DELL 9 O ' 'GBAGCbI DATE 5/ 3 /97 5/8/97 5/T/97 / /97 / /97 0FF:CIAL RECORD COPY a l l
The Files i EPRI presented a process to categorize the consequences of piping segment l l failures based primarily on a qualitative analysis of plant specific success paths for mitigating accident sequences. EPRI stated that it was possible to l perform the analysis under the quality assurance guidelines applied to other l deterministic analyses used for licensing' purposes. A plaat specific Probabilistic Risk Assessment (PRA) could be used to provide quantitative support to the consequence categorization, but EPRI envisions PRA quantification to be used generically to calibrate the process, and to resolve complex interactions in specific instances. I Industry representatives also provided an executive update of pilot studies for Fitzpatrick and ANO-2 plants. Industry representatives stated that utility submittals for pilot plants are scheduled for September 1997. l NRC staff commented that the performance monitoring feedback needs to be l included in the EPRI RI-ISI process. NRC staff also indicated that the RI-ISI process must address risk as discussed in the Draft Regulatory Guide DG-1061. i The staff members provided additional comments on the details of the EPRI j l methodology which will be documented in the Request of Additional Information j (RAI) to be transmitted to NEI/EPRI. The meeting concluded with the agreement that additional public meetings will be held to discuss further details of EPRI methodology and its application to pilot plants. i Attachments: As stated. l l l I i l 1 i
l The Files May 12, 1997 l \\ EPRI presented a process to categorize the consequences of piping segment ) failures based primarily on a qualitative analysis of plant specific success paths for mitigating accident sequences. EPRI stated that it was possible to i perform the analysis under the quality assurance guidelines applied to other deterministic analyses used for licensing purposes. A plant specific Probabilistic Risk Assessment (PRA) could be used to provide quantitative ) support'to the consequence categorization, but EPRI envisions PRA quantification to be used generically to calibrate the process, and to resolve complex interactions in specific instances. Industry representatives also provided an executive update of pilot studies for Fitzpatrick and ANO-2 plants. Industry representatives stated that utility submittals for pilot plants are scheduled for September 1997. l NRC staff commented that the performance monitoring feedback needs to be included in the EPRI RI-ISI process. NRC staff also indicated that the RI-ISI process must address risk as discussed in the Draft Regulatory Guide DG-1061. l The staff members provided additional comments on the details of the EPRI l methodology which will be documented in the Request of Additional Information (RAI) to be transmitted to NEI/EPRI. The meeting concluded with the agreement that additional public meetings will be held to discuss further details of EPRI methodology and its application to l pilot plants. 1 j Attachments: As stated. DISTRIBUTION: File Center (T-5 C3) PDR (LL-6) ECGB RF NRR Mailroom NRC Meeting Participants AThadani GHolahan MMayfield JJohnson GMillman Glainas JMurphy WHodges LShao JStrosnider KWichman RJones JAustin RWessman DFischer i DISK / DOCUMENT NAME: G:\\ALI\\ SUM 04978 To receive a copy of this document, indicate in the box: "C" = copy w/o attachment; "E" = Copy w/ attachment; "N" s Noj L OFC ECGB:DE ECGB:DE ECGB:DE NAME SALI* P0 DELL
- GBAGCHI*
l DATE 5/8/97 5/8/97 5/8/97 / /97 / /97 l OFF,CIAL RECORD COPY
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i r - EPRI EPRI-NRC TechnicalMeeting t EPRI i RISK-INFORMED INSERVICE INSPECTION H presentedby: 4 l b S. R. Gosselin, P.E. o! Electric PowerResearchInstitute r i 3' (T\\ Jim Chapman t i Vesna Dimitrijevic dl Patrick O'Regan go l Yankee Atomic Electric Company KarlFleming j ERIN Engineering & Research April 29,1997 t Component integrity Technology a SG042997. PPT - 1 d I
- EPRI i l Summary \\ i INTRODUCTION - EPRIPROCESS - PILOT STATUS 1 PSA BASIS AND APPLICATION l FAILURE POTENTIAL via DEGRADATION ASSESSMENTS t RISK IMPACT ASSEMENTS i QUALITATIVE QUANTITATIVE l CONLUSION i Component Integrity Technology a SOO42997. PPT - 2
4 F - EPRI I KEY POINTS Demonstrate quantitative basis of the consequence analysis j Deterministic principles are addressed and defense-in-depth is maintained Process driven approach that yields consistent results Qualitative analysis of risk impacts are wen understood and manageable e Confirmatory research to address risk impacts of RI-ISI process i i Component integrity Technology m__ i I
I t I - EPRI l I L i i i f r i i Ilf i INTRODUCTION t t t t S. R. Gosselin, P.E. Electric Power Research institute l I L l i Component Integrity Technology f m_ f SOO42997. PPT - 4 { t
i I i m 1 \\ l Risk-informed ISI Process l 1 i SYSTEM SCOPE l All Cless 1,2, & 3 Piping l Non-Code Piping Important to Safety + r FAILURE MODE & EFFECTS i j ANALYSIS l CONSEQUENCE l ASSESSMENT I Direct & Indrect Effects Deterministic insights e l } FAILURE MODE ASSESSMENT Design & Operating Condtions industry Service Experience 4 j-Degradation Mechanisme Other Deterministic insights t j ( PlPE SEOMENT DEFINITION ) t-2 i + t r j RISK EVALUATION Segment Consequence impact Assessment Core Damage l Containment integrity t e m Segment Failure Ukelihood Assessment v l r 3 Segment Risk Categorization e + PLANT EXPERT PANEL integreed Plant Rwiew Operational Experience Other Deterministic insights t + SAFETY SIGNIFICANT SEGMENTS ComponentIntegrityTechnology =
.I i i = 3 4 i Risk-Informed ISI Process l (continued) 4 i 1, j r SAFETY SIGNIFICANT SEGMENTS h i r I SEGMENT INSPECTION LOCATION SELECTION j Procedural Criteria & Service Experience l t i 4 l EXAMINATION l REQUIREMENTS j Apply Prescriptive Requirements: 1 Examination Method Examination Volume j Acceptance Criteria l Evaluation Standard l I j r i t Document Revised ISI Plan i i ( i i i ] t } ) ComponentIntegrityTechnology = i. ROCESS. PPT 2 i ( ) b ~
t - EPRI i I f RISK CONCEPT RISK = (Core Melt Potential l Pipe Rupture) x (Pipe Rupture Frequency) Conditional Core Melt Potential Pipe Rupture Frequency j Degradation Mechanisms j PSA & Deterministic Service Experience Impact Groups - Initiating Event Rupture Potential Ranking l - Degraded Containment HIGH - Degraded System /frain - MEDIUM Combination LOW Consequence Ranking - HIGH - MEDIUM LOW Component Integrity Technology f w 1 i
EPRI R SK MATR X i: CONSEQUENCE CATEGORY = Core Melt Potential for Umiting BreEk'SIAe? RISK REGIONS ~. HIGH MEDIUM LOW NONE LQE MEDIUM HlGli . ~,,, e, c.. ,, -.~ $', -,U, < 's (' )x 4' y,. x,' h " 2, ' ", ~
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t - EPRI I s Risk-Informed ISI Evaluation Procedure Technical Reports l I EPRI Report TR-106706, " Risk-Informed Inservice l Inspection Evaluation Procedure - Interim Report", l June 1996 ERIN Engineering, " Independent Review EPRI Risk-f Informed inservice Inspection Procedure, Final Report", September 1996. j i Structural Integrity Associates, " Review of Degradation Mechanisms in EPRI Risk-Informed Inservice inspection Evaluation Procedure", SIR 097, November 1996. I i l Component integrity Technology l er SG042997. PPT - 9 l
- EPRI i Supporting Technical Papers l t Chapman, et al, "A Practical Approach to Risk-Based inservice inspection in U.S. j Nuclear Power Plants", PVP Vol. 339, Pressure Vessels and Piping Codes and j Standards, Volume 2, ASME, N.Y.,1996. l Gamble, R. M. and Gosselin S. R., "A Method to Define Degradation Mechanisms. and Failure Rates for Piping", PVP Vol. 339, Pressure Vessels and Piping Codes and Standards, Volume 2, ASME, N.Y.,1996. f Gosselin, S. R. and Fleming, K. N., " Evaluation of Pipe Failure Potential via i Degradation Mechanism Assessment", Proceedings of 5th International Conference l on Nuclear Engineering, ICONE 5, ASME, N.Y.,1997. Dimitrijevic, V. B., O'Regan, P.T., and Smith. A, " Risk-Based ISI Application to a Boiling Water Reactor", PVP Vol. 339, Pressure Vessels and Piping Codes and Standards, Volume 2, ASME, N.Y.,1996. O'Regan, P. J. and Dimitrijevic, V. B., " Risk informed Inservice Inspection insights from a BWR Pilot Evaluation", Prceedings of 5th International Conference on Nuclear Engineering, ICONE 5, ASME, N.Y.,1997. O'Regan, P. a. and Dimitrijevic, V. B., " Insights from Risk informed inservice j inspection of a BWR Pilot Plant", to be published in 1997 ASME PVP Conference Proceedings. f i Component integrity Technology w SOO42997. PPT - 10 [ i
l - EPRI l BWR Pilot Application Study I I b EPRI T/C Project with NYPA Pilot Plant: James A. Fitzpatrick (GE-BWR/4) Scope 14 systems All Class 1, 2, & 3 piping and non-code piping important to safety Status Risk evaluation complete on all systems Integrated plant reviews ongoing l Schedule f Utility submittal forecast September 1997 i t l Component integrity Technology l sooam.wr - si i
I l I EPRI i i I r J. A. Fitzpatrick Systems i t i SYSTEM TITLE CODE Ci ASS 1 Reactor Water Recirculation RWRS 1 2 Main Steam MS 1 3 Main Feedwater MF 1 [ 4 Core Spray CS 1, 2 l 5 Reator Water Cleanup RWCS 1 6 Control Rod Drive CRD 2 7 High Pressure Coolant injection HPCI 1, 2, NNS 8 Residual Heat Removal RHR 1, 2 9 Reactor Core Isolation Cooling RCIC 1, 2, 3, NNS l 10 Nuclear Boiler Vessel Instrumentation INST 1 r 11 Standby Uquid Control SLC 1, 2 l 12 Fuel Pool Cooling FPC 3 l 13 Service Water & RHR Service Water RHRSW 3,NNS t 14 Emeraency Service Water ESW 3 i i Component IntegrityTechnology SOO42997. PPT - 12 1 I
EPRI CE-PWR Pilot Application Study i r EPRI T/C Project with 6 CE Utilities f Pilot Plant - ANO Unit 2 f Scope 10 systems (RCS, HPSI, LPSI, SDC, CS, CVCS, MS, MFW, AFW, SW) a All Class 1, 2, & 3 piping and non-code piping important to safety i Status Risk evaluation completed on all systems l Plant integrated review of risk evaluation completed Inspection location and examination selections in progress Schedule Utility submittal to NRC for all system (w/e service water) June 1997 i Utility submittal to NRC for service water system September 1997 i i l t Component integrity Technology m_ 4 somm. PPT - 13 i
- EPRI ANO SYSTEMS 1 SYSTEM TITLE CODE CLASS 1 Reactor Coolant System RCS 1 i 2 Chemical Volume & Control CVCS 1, 2 3 High Pressure Safety injection HPSI 1, 2 4 Low Pressure Safety injection LPSI 1, 2 5 Shutdown Cooling SDC 1, 2 f 6 Containment Spray CS 2 7 Main Steam MS 2 I 8 Main Feedwater MF 2, 3, NNS 9 Emergency Feedwater EFW 2, 3, NNS i 10 Service Water SW 2, 3. NNS i I i t l l [ t Component integrity Technology su SG042997. PPT - 14 (
4 Technical Basis of Approach Demonstrate Sound Technical Basis - Probabilistic Basis - Deterministic Consideration i s Establish Quantitative Risk Basis of Approach j - Potential for Pipe Breaks sl - Impact.of Breaks on Risk 3! Ability to Forecast Impacts of Changes to ISI Program i f l
'%-r'* Risk Assessment Challenge Compare the change in relative risk given a change'in ISI in full consideration of: 1. Existing PSA models and methods 2. Service Experience (i.e., new information per SECY-95-280) 3. PBF uncertainties 4. ISI influence 5. Existing programs (e.g., E/C, MIC, IGSCC) I i 6. Redundancy and defense-in-depth (i.e., deterministic per SECY-95-280) t 7. Level of discrimination (i.e., no. of segments) 8, Expert panel involvement Based on a review and evaluation of alternative approaches, we have strived to develop a " Risk Informed" process that meets this challenge. j i l
l s-r I Development of " Risk-Informed" Process - i i 1. Considered 1 Typical PSA ranking and change approaches combined with l PBF models and experience Other industries (insurance, oil, gas, etc.) Fundamental PSA framework (i.e., back to underpinnings of PSA) 2. Concluded Uncertainties were dominant challenge, especially among systems and within individual pipe segments Back to the underpinnings of PSA required l t i i
Development of " Risk-Informed" Process (Continued) 3. Uncertainties could best be addressed by semi-quantitative j systems-oriented approach that: 1st focused on individual systems to assess risk ranking and change impact among elements within the system vs. comparing elements across all systems l 2nd ensured all potential functions, modes, and failure times j were explicitly analyzed vs. relying on expert panels as j an explicit input 3rd used a progressive screening and focusing approach j (many risk bins to 12 to 3) 4th ensured risk matrix (i.e., ranking categories) addressed j both deterministic and PSA measures j i
~. 1 ? Development of " Risk-Informed" Process (Continued) l 5th ensured the final 3 risk categories address precursor j events, high consequence events, even if no damage j mechanisms were identified, and i 6th ensured likelihood and consequence uncertainties deterministically and probabilistically j i l A GRADED, PROGRESSIVE APPROACH with additional quantification and comparison among systems as an option i
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I* [ i Grouping h t Risk categories are combined into three to: l l, i ~! 1. Accommodate uncertainties i i 2. Facilitate establishing an ISI sampling percentage j i 3. To ensure high consequence segments without identified damage mechanisms are considered 4. To ensure segments with the potential for large leaks (implying relatively high failure frequencies) are considered j even if the consequence category is low t l
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Deterministic Principles for Consequences I I l t 1. Initiating Events: j Design basis event Category A adjusted for PSA 2. Mitigating Systems Redundancy (backup trains) ~ i i Exposure r t i 3. Exposure Test t Standby i i Demand Operation j 4. Combinations i PRt.HAPMAN\\$HiMlbl\\22
yy '.* : S EPRI Risk Impact ( Consequence) Categories Severe initiating Events l High-risk Regions CCDP>10-4 Severe Loss of Mitigation { High Risk of Bypass l Moderate Initiators I Medium-risk Regions 10-6 <CCDP<10-4 Moderate Loss of Mitigation Moderate Risk of Bypass l Mild initiators Low-risk Regions CCDP<10-6 Minimal Loss of Mitigation l Full Containment .mj TW
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t t Summary of Evaluation Process 5 1. Evaluation at system level 2. Use a qualitative, semi-quantitative risk analysis to group segments within a system into three inspection risk regions (H, M, L) j i F l 3. Address uncertainties and " precursor" events i 4. Accommodate " full" quantification as an option l 4 l F4hlHAFMAPIN(!W51\\3 7 i
t i Conclusion l 1. Existing PSA methods and knowledge make it difficult to fully quantify the impact of ISI changes across all systems 2. Performance history, fundamental PSA practices, PSA results, and deterministic analysis are combined in our process i l PH\\LHAHAArf64tSuhl\\39 - ---- ------ L
i t Conclusion (Continued) l l, t 3. Performance history is considered at 3 frequency levels and in l establishing ISI changes i l 4. Fundamental PSA practices are used, including both qualitative and quantitative measures, including 4 severity categories i e I r l m s.vu,v mususn.e i i f i \\
i Conclusion (Continued) i i l i i t l 5. PSA results are used, including CCDP, full spectrum of IEs, all safety functions, spatial effects, and containment performance j i i 6. Deterministic considerations include design basis events, SF criteria, redundancy, spatial effects, and coupling of RCS and containment; precursor events are addressed ~ { emwouums.ows r r
i l l~ t Conclusion (Continued) i i 7. Results are insensitive to number segments j t i 8. Process uses peer review for additional assurance of quality l I ? ) i i f mm.wwwu2 r Ii i I
I t l I l I PSA Principles for Consequences ? 1 L Analyzed Effect of Measure Used to l Configuration Pipe Failure Rank Consequences I 1 i I f 1. Operation, Initiating Event Conditional Core Damage Probability -4 [ (Test, Standby) Given the Initiating Event Nf n I 3 L m Conditional Core Damage Probability t. j d i 2. Test, Standby, Loss
- of Given Loss of System / Train and Operation System / Train Exposure Time Equal to Detection Time l
Plus Applicable AOT l I Conditional Core Damage Probability i 3. Demand Loss
- of Given Loss of System / Train and System / Train Exposure Time Equal to Time Between Tests or All Year l
Loss
- refers to both: total loss or degradation.
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l 1 i l f l ? Consequence Categories for Pipe Failures Resulting in initiating Event (JAF Pilot Plant) l i r I I Recommended i Design Basic Events Initiating Event Consequence CCDP j Category Type Category L T3A - Transient with Condenser Available Low <1 E-6 II Anticipated Occurrance T2 - Loss of PCS Low <1 E-6 f T3B - Loss of FW with Condenser Available Low <1E-6 I T3C - IORV Low 1 E-6 Infrequence TDC - Loss of Safety DC Bus Low <1 E-6 Events TAC - Loss of Safety AC Bus Medium 4 E-5 T1 - Loss of Off-Site Power
- High
>1 E-4 S3 - Small Small LOCA Low <1 E-6 i S2 - Small LOCA Low <1 E-6 i IV Accidents S1 -Intermediate LOCA Medium 3E-6 l a A - Large LOCA Medium 7E-5
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t l JAF Initiating Event Frequency, Conditional Core Damage Probability, and EPRI Consequence Category JAF Initiating Events and Conditional Core Damage Probability (CCDP) Consequence Assignment (Table 2-1) Initiator Event Events /yr CCDP CCDP Comments Category Explanation (in addition to CCDP) T1-LOSP 0.057 3E-5 based on 1.8E-6 High Moved from Medium to High (>lE-4 CDF in Table 3-3 because itis assumed that LOSP is without not as easily recovered ifit is caused recovery) by a pipe break T2 - loss of PCS 0.593 <1 E-7 2E-7 if all of 1.2E-7 Low in Table 3-3 is used T3A - PCS available 4.72 <l E-7 based on Table 3-3 Low Figure 3-1 T3B - loss of FW 0.445 <1 E-7 3E-7 if all of 1.2E-7 Low inTable 3-3 is used T3C -IORV 0.102 <1 E-7 based on Table 3-3 Low Figure 3-1 L -large LOCA 1E-4 7.4E-5 assumes all of 7.4E-Medium low pressure permissive dominates 9 in Table 3-3 common cause oflow pressure ECCS and no credit for manual actions. SI -intermediate LOCA 3E-4 3E-6 2E-5 if all of 7.4E-9 Medium Figure 3-1, dominated by high in Table 3-3 is used pressure makeup assuming RCIC is inadequate and the MSIVs close S2 - small LOCA 3E-3 <1 E-6 based on Table 3-3 Low Figure 3-1 and success criteria S3 - small small LOCA 3E-2 <1E-6 based on Table 3-3 Low Figure 3-1 and success criteria TAC -loss of safety AC bus 5E-3 4E-5 6E-6 if all of 3.2E-8 Medium in Table 3-3 is used TDC -loss of safety DC bus 5E-3 <1 E-6 based on Table 3-3 Low
Consequence Category Assignment for ANO-2 Pipe Failures Resulting in Initiating Events Initiating Event IEF CDF CCDP Consequence T1 - Turbine trip 7.6E-01 2.3E-06 3.0E-06 medium T2 - Loss of PCS 2.5 E-01 9.0E-07 3.6E-06 medium T3 - LOSP 5.8E-02 1.7E-06 2.9E-05 medium T4 - Excessive FW 9.4E-04 1.9E-09 2.0E-06 medium T5 - Steam / Feed break 1.1E-03 1.1E-09 1.0E-06 medium T6 - Reactor Trip 2.0E+00 6.0E-06 3.0E-06 medium T7 - Loss of SW 5.5E-03 2.1E-06 3.8E-04 high T8 - Loss of SW P4A 7.4E-02 2.1E-07 2.8E-06 medium T9 - Loss of SW P4B 7.4E-02 2.0E-07 2.7E-06 medium T10 - Loss of DC D01 3.9E-04 9.8E-06 2.5E-02 high Tl1 - Loss of DC D02 3.9E-04 1.1 E-06 2.8E-03 high T12 - Loss of AC A3 3.9E-04 3.2E-06 8.2E-03 high T13 - Loss of AC A4 3.9E-04 5.8E-08 1.5E-04 high T14 - Loss of AC B5 1.0E-03 1.9E-07 1.9E-04 high T15 - Loss of AC B6 1.0E-03 1.2E-07 1.2E-04 Ingh S - Small LOCA 5.0E-03 1.7E-06 3.4E-04 high M - Medium LOCA 1.0E-03 1.7E-06 1.7E-03 high A - Large LOCA 1.0E-04 1.4E-06 1.4E-02 high R-SGTR 9.8 E-03 9.5E-08 9.7E-06 medium ISLOCA (1) (1) (1) high 1 cf.( 3) = 0((2Y0?[(t) i Draft 1/8/97
l l t Consequence Categories for Pipe Failures Resulting in System / Train Loss Affected Systems l Number of Unaffected Backup Trains l Frequency of Exposure Time t 0 1 2 23 Challenge Challenge www a I hhMyd'g.k lj L'j$,, l All year l Between tests Anticipated M ! L. (1-3 months) Long AOT Na[p[m --'$7 -g l (DB Cat II) (1-2 weelcs) d?$ 4 $y$ N N Y h @h. E ~- Short AOT(s3 days) M ' i-Ch L'- All year M
- C Infrequent
- rs -
_c .fng;rr = NS mA Between tests / kh[4h $h ~ (DB Cat 111) (1-3 months) 1 - 3 Long AOT M L. ~j L (1-2 weeks) w */IN?N u [ery-m w. S L Short AOT(s3 days) h kh - (L ' All year Unexpected Between tests 4 M
- " L' -
L ~' (DB Cat IV) (1-3 months) Lon9 AOT D N Y* N ID ~ =4 u i , ' L' ' [ (1-2 weeks) y@% gi 22 L ~ L+~ ~ ~ t Short AOT(<3 days) ^L: A , LJ 'L^ I i i I PROMITmIRSIAPP
Consequence Categories for l l Pipe Failures Resulting in System / Train Loss i i 4 Affected Systems l Number of Unaffected Backup Trains l Frequency of Exposure Time to 7 i 2 2 Iv 23 0 1 m Challenge Challenge 3 l a v ~.- _.w m% e lgNd. iGEj06h All year I jgE i Between tests ) Anticipated 4 2.5E-05 2.5 E-7. - (1-3 months) 11.9E-d8 ' ' (DB Cat il) l (1-2 weeks) 52
- gF: h qwgg I
nyy s.. Short AOT(s3 days
- 2.7E-05 12.7 E-07-;
- 2.7 E-09.
j i All year 1.0E-05~ (1.0E-07? l Infrequent gggpA - J.5E-08 I Between tests t.SEM!f$ 2 (1-3 months) yn aggg;ye (DB Cat 111) ' 1.9 E-05 1.bE-07E .f1.9 E I / ong AOT J0 (1-2 weeks) w j ., n,, m.n. [ Short AOT(s3 da .7E-06Mi'
- .'.l2.7E-081,
f 2.7E-10 l ^ i [ (1.0E-08 : All year t ^ Unexpected m 1 $E-Oh. 2.5E-092 Between tests 2 2.5E-05 (DB Cat IV) (1-3 months) j Md 1 Long AOT ~.9 b-10 $3$ 1.9h-08:. I (1-2 weeks) .u 4 Short AOT (<3 days) / 2'.'7 E O7 i 2.7E-11 I
- 2.7 E-09 I
PRIHMITRU\\RIStAPP l ]
t JAF Mitigating Functions high pressure low pressure heat removal l makeup makeup I RCIC LPCI"A" RHF1 "A" i ~ I i t HPCI LPCI "B" RHR "B" l initiating -+ Success -+ Event Core t Main Cond. PCS -+ Spray "A" Venting 2 of11 SRVs -+ Sp ay B" t I i Condensate i 4 RHRSW i I i ? PfhDtM'TRURSIAPP h
i Assumed System & Train Backup l System / function / train (1)(2) Trains Unavail Explanation l HPSI Inj train (independent) 1 1.2E-2 Split fraction HAl in PRA Table 8.2.1.5 i LIPSI Inj train (dependent) 0.5 0.19 Split fraction HB2 in PRA Table 8.2.1.5 HPSI Recirc train (independent) 1 7.5E-3 Split fraction CA1 in PRA Table 8.2.1.5 l HPSI Recirc train (dependent) 0.5 7.2E-2 Split fraction CB2 in PRA Table 8.2.1.5 j LPSIInj train (independent)s 1 3.5E-3 Split fraction LAl in PRA Table 8.2.23 LPSIInj train (dependent) 0.5 0.14 Split fraction LD1 in PRA Table 8.2.23 [ Spray supply valve 1 1.2E-2 Split fraction val in PRA Table 8.2.43 l (independent) Spray supply valve (dependent) 0.5 6.4E-2 Split fraction VB3 in PRA Table 8.2.43 l Spray inj train (independent) 1 6.4 6 3 Split fraction ASI in PRA Table 8.2.43 Spray inj train (dependent) 0.5 8.8 E-2 Split fraction BS2 in PRA Table 8.2.43 Spray Recirc(independent) 1 9.2E-3 Split fraction YAl in PRA Table 8.2.4.3 Spray Recire (dependent) 0.5 0.15 Split fraction YB3 in PRA Table 8.2.43 l RHR HE(independent) 1 73E-3 Split fraction XAl in PRA Table 8.2.43 RHR HE (dependent) 0.5 7.662 Split fraction XB5 in PRA Table 8.2.43 l SI Actuation (independent) 1.5 1.0E-3 Split fraction sal in PRA Table 8.2.6.1.1 j' SI Actuation (dependent) 0.5 6.5E-2 Split fraction SB2 in PRA Table 8.2.6.1.1 Recire Actuation (independent) 1.5 1.063 Split fraction RAI in PRA Table 8.2.6.2.1 Recirc Actuation (dependent) 0.5 73E-2 Split fraction RB2 in PRA Table 8.2.6.2.1 Main feedwater/oper success 1 3.963 Split frac. tion MF1 in PRA Table 8.1.2.4 i Main feedwater/oper fails 1 3.2E-2 Split fraction MF2 in PRA Table 8.1.2.4 - l MFW/AFW/EFW/oper success 3 1.6E-7 Split fraction SHLIS in PRA Table 8.1.2.5 l MFW/AFW/EFW/oper fails 2.5 1.4E-5 Split fraction SHL2F in PRA Table 8.1.2.5 l l l l Draft 4/19/97
JAF IPE Success Criteria for Transients & Small LOCA With PCS Available high press makeup low Press makeup heat removal l "A" MR "A" RCIC (0.1) (1L2) (1L2) B" N B" HPCI(0'1) (152) (IL2) Initiating - -+ Success Event (PCS) Core Spray *A" Main Cond .y (IL2) (162) (IL2) i 2 of 11 SRV Core Spray "B" Venting (1L2) (IL3) (IL2) t i ConderJate [ (IL2) [ t Other(1E-2) (RHRS%) I i i I f f l . =
I Core Spray System Train A f i 1 (cs-c-5A ) (cs-c-4 A) ( CS-C-3 A) i h (RB-300Al (Rg.300A3 [ tRB-30048 M V.m f MOV11A MOV12A { ~ ( CS-C-10A) - tRB-272Al h 22 e mm 1 r:T (C5-C-6 A) Mov26A i(CS-C-1 A) [ (RB-2270) [ f a !8! ~ V i (cs-c-24): RO27A idOv5A I W i { Aavi3A 14,A i4e Aaviss . c i (RB-2270) I ~ ( -v: 2 Failures "None" j i i Note: Active failure is " failure to close" on demand. l Passive failure is " failure to remain closed." meaumuese
i Core Spray f l Possible Consequences & Rank Consequence Rank CCDP 1. LOCA Medium 7 E-5 2. ISLOCA High -:. 5 E-3 3. Isolable Loss of Torus and ECCS Low SE-7 4. Unisolable Loss of Torus and ECCS ' Medium S E-6 5. Loss of CS One Train Low <1 E-7 u
.s as-pag Amsma6 M4e-,, m o am 2a&Kw&AuLJm-=-naa mm nm am-M "--3-m&AJs.A,eM&-0mM.mMM eMAGMe Le M M4 MA HM+mM4 s *M-mM4md*' Ae AM-E a =aK& MaAm4 4e4e & B4 9E+A4~&S JA9A 4=a4 A 4 = A i I i i C l O N O I O i Q. E l 4 E_ a en x-LU E 1 i M } I L i LU j i I Il i I i t A
Basic Tasks
- 1. Identification of the piping system configuration and scope of the analysis
- 2. Identification of consequences 4
i f
- 3. Identification of applicable degradation mechanism (s?
i 1
- 4. Assessment of risk importance
- 5. Selection of inspection locations
- 6. Selection of appropriate inspection techniques i
- 7. Documentation of the revised Section XI Program l
i mourmunise
1 i JAF Mitigating Functions 1 i high pressure low pressure heat removal makeup makeup i RCIC LPCI "A" RHR "A" i t HPCI LPCI "B" RHR "B" Initiating _* Event -+ Success ~ ~ Core i PCS Spray "A" Main Cond. t Core 2 of11 SRVs Spray "B., Venting j ~ Condensate I l i RHRSW
I Core Spray System t i Two independent trains Each train provides 100% low pressure makeup capacity l Auto actuates in response to a loss-of-coolant-accident i ILOCA) Suction: Suppression Pool 1 Torus) i One normally open motor-operated valve j I:MOV) is located between Torus and pump Discharge: Reactor Coolant System i:RCS) One normally open MOV and one normally closed MOV are located between pump and RCS mourmuniswe
1 t [ ~ I t t Core Spray System Train A i i t I (Cs-C-5 A) (Cs-C-4A) ( Cs-C-3 A) (RS-3 COA) (RB-300A3 fRS-300A) )N WOV12A l uoV11A i I i ( CS-C-10A) _ epa.272An E 3 h IEE:$$?i! O. I mm - e !(CS-C-IA) l (C5-C-6A) 6A (R8-227DP i j } !I! [CS-C-5) RO27A MOVSA 1 M N LO A L0 Lo i " *:\\ i ADV13A 14A 148 ACV138 . II 0 l (R8-22703 I M v. m i h woY,7A f '^ (~CS-C-? A ) f (CS-C-8A) f (CS-C-9A). ((RS-2270) } (R8-227Al I (R8-227AlL l l t i l J f i -, __n [ l
Affected Areas t i Torus Room (RB 227A) Propagates into West and East Crescent areas l
- West Crescent Area (RB 227D)- Train A j
Propagates into Torus Room and East Crescent area i i
- East Crescent Area (RB 227B)- Train B l
Propagates into Torus Room and West Crescent area t t
- Reactor Building, Elevation 300' (RB 300) l Propagates to both Crescent areas k
k
i i Type of Consequences t i LOCA inside Drywell j o t Increased potential for a LOCA outside of Drywell (ISLOCA) o isolable loss of suppression pool inventory outside of o Drywell, combined with f!4ooding of all makeup systems (ECCS) i i i Unisolable loss of suppression pool inventory outside of o Drywell, combined with flooding of all makeup systems (ECCS) I l PROMITRLARfSAPP 1
-..___..._..._._..___._.1 l Consequence: CS-1 A Location: Between reactor and Check Valve AOV 13A in Drywell j i Initiating Event: LOCA in Drywell l Isolability: No I Loss of System: No l; Loss of Train: CS-A Description of The worst case intermediate LOCA i Consequences (MLOCA)is assumed. RCIC does not provide adequate makeup and MSIVs will close on RPV Low Level causing loss of PCS. l Rank: MEDIUM i Basis for Rank: Initiating Event impact: MLOCA is DB event Category IV. LOCAs are not significant contributors to JAF j plant risk. j 1 j moumuuuse i'
30-Sep-96 FMECA - Consequence Information Report Consequence ID: CS-C-Ol A Pipe Section
Description:
Between RPV and CV AOV 13A , Spatial Affects: Containment Affected Location: Drywell Spallal Effects Comments: No different than design basis LOCA. Flow diversion pumps torus to drywell which then returra to torus. Initiating Event: LOCA Initiating Event Isolability: No, piping is between reactor vessel and first containment isolation valve. PotentialInitiating Event: PotentialIE Prevention: Configuration: Standby (Initiating Event) DirectImpact: CS A (flow diversion orisolation) Indirect Impact: None Avall. Means of Isolation: Pump trip and MOVs, but irrelevant to consequences. ) Avail. Trains ofIIP Makeup: Unaffected Avail. Trains of LP Makeup: 2 LPCI + CS B = 2 trains kvallTrains ofIIcat Removul: Unaffected Containment Performance: Unaffected Other Operating Modes: CS also provides standby reactor makeup function during shutdown. When the reactor is shutdown & depressurized, this piping is not pressurized and is less likely to fail, and there is more time and additional makeup sources. Importance is judged not to increase. External Events: The pipir g failure is assumed during operation / standby when it results in an initiaGng event. External initiating events are not considered in this case. Consequence Category: MEDIUM Consequence Comments: LOCA initiator results in a " MEDIUM" consequence rank (Table 2-1). l t + l i 4
Consequence: CS-3A j i Location: Upstream of CV AOV 13A in RB300A Initiating Event: Potential ISOLOCA l Isolability: Yes, Check Valve AOV 13 l { Loss of System: No Loss of Train: CS-A Description of LOCA outside drywell can occur if one j Consequences: passive failure occurs: failure of Check Valve AOV 13A. Rank: HIGH l Basis for Rank: Increased potential for ISOLOCA impact: j only 1 passive protection PRDnA!TRLfAISIAPP
O j 30-Sep-96 FMECA - Consequence Information Report l Consequence ID: CS-C-93A l Pipe Section Descripuou: Upstream of CV AOV 13A outside Drywell. i i Spatial Affects: Propagation Affected Location: RB300A j l Spatial Effects Comments: Potential ISLOCA (CV failure) is assumed to pr.opagate steam throughout the Reactor Building failing PCS (MSIV closure due to RB environment or reactor level) and ECCS. Water from CS flow diversion drains the torus with propagation to RB227B & RB227D (potential loss of all ECCS). l Initiating Event: I initiating Event Isolability: PotentialInitiating Event: ISLOCA in Reactor Building if a passive failure of CV AOV 13A occurs during operation. PotentialIE Prevention: Norraally closed CV AOV 13A Configuration: Standby or Demand (no periodic testing). Demand is evaluated as worst case (MLOCA assumed). Direct Impact: CS A (flow diversion or isolation). Indirect Impact: All ECCS due to flooding and draining torus Avail. Means of Isolation: Pump trip and MOVs. Flow diversion can be isolated. Avail. Trains ofIIP Makeup: Not applicable because CS demand occurs after HP Makeup success (MOV 12A opens on low pressure permissive). Avall. Trains of LP Makeup: RHRSW = 1 train kvallTrains ofIIcat Removal: Vent = 1 train Containment Performance: CV AOV 13A Other Operating Modes: CS also provides standby reactor makeup function during shutdown; there is more time and additional makeup sources when the reactor is shutdown & depressurized. Importance is judged not to increase. External Events: For seismic, one can assume the seismic capacity of pipe is high and the frequency of challenge is comparable or less than the frequency used in the evaluation. Similarly for fires, the frequency of challenge should be comparable or less than the frequency used in the evaluation. There are no external events which significantly degrade the low pressure makeup function. Importance is judged not - to increase. Consequence Category: HIGH Consequence Comments: Potential ISLOCA results in a "HIGH" consequence rank based on Table 2-4 (1 passive check valve failure). Demand impact results ir. a " LOW" consequence e l rank based on Table 2-2, assuming an unexpected frequency of challenge, all year exposure time and 2 backup trains: RHRSW and 2 means of isolation. Two l means to isolate are credited as 1 train because of the high stress situation. l l 8 1
Consequence: CS-9A l Location: Between Suppression Pool and MOV 7A l at Torus Room i Initiating Event: No, potential manual shutdown due to e loss of Torus level Isolability: No Loss of System: Possible loss of all ECCS systems if j Suppression Pool is lost j Loss of Train: Description of Unisolable loss of Suppression Pool Consequences: inventory. Torus water level alarm and l floor sump alarm will provide operator with information about event. Rank: MEDIUM Basis for Rank: . System / Train impact: Unexpected frequency of challenge, between test exposure time (3 months), one backup train (RHRSW). Containment bypass. me-umue-
CS A Consequence Eva uation Results Consequence Pye Section eW Initmating test Awastaldo Contamment Exposure Table ID Description t.ncation Co";-M-= Event (IE) Isolaemon Trains Trains Effects Tune Used Rank CS-C-01 A Between RPV & CV Drywel Standby LOCA No LOCA effects Unaffected N/A 2-1 MEDIUM AOV 13A [Lbkl CS-C42A Upstream d CV AOV Drywel Standby Pdental CV AOV 13A LOCA effects 2MOV& N/A 2-1 13A hside Drywet LOCA Closed System
- LONi Demand Assumed:
Pump & MOV CS-A 2:CS-B & 2MOV& 1 Year 2-2 Medum LPCI A/B Closed System
- u.-
LOCA [Hidii CS-C-03A Upstream d CV AOV RB300A Standby Paental CV AOV 13A Al CV AOV 13A None NA 2-4 13A Outside Drywel ISLOCA ~ k M,y] Demand Assumed: 1 Pmp & ECCS 1:RHRSW CA AOV13A 1 Year 2-2 ELOM/ij Medum MOV 'tj LOCA CS-C44A Between MOV 11 A & RB300A Standby Potental CV AOV 13A AI None CV AOV 13A N/A 2-4 [ LOW MOV 12A ISLOCA & MOV 12A & MOV 12A ~ Dernand Assumed: 1 Pump & ECCS 1:RHRSW CV AOV 13A 1 year 2-2
- LbW F
Medum MOV & MOV 12A LOCA L CS-C-05A Upstream of MOV RB300A Dern:.' i As ;umed: 1 Pump & ECCS 1:RHRSW MOV 7A 1 to 3 2-2 LOW : 11 A at EL 300 Medum MOV Morths LOCA CS-C-06A Upstream d MOV RB272A Demand Assumed: 1MOV 7A ECCS 1:I' HRS W MOV 7A 1 to 3 2-2 LOW.. 5 11A at EL 272 & 244, RB227D Medum Morths ~; : above Torus level LOCA CS-C-07A Ptrnp Suction & RB227D Demand Assumed: 1MOV 7A ECCS 1:RHRSW MOV 7A 1 to 3 2-2
- LOW >
l Discharge bekw Medum Morths
- li Torus level LOCA
/g CS-C-08A Isolable Pump RB227A Demand Assumed: 1/2:MOV7A ECCS 1:RHRSW hOV 7A t to 3 2-2 MEDIUM Sucion in Torus Medum Morths i Room LOCA CS-C-09A Ptsnp Suction RB227A Demand Assumed: No ECCS 1:RHRSW Bypass I to 3 2-2 MEDIUM between Torus and Medum Morths MOV7A LOCA [LdW)! CS-C-10A Test Retum Line RB227A Test No 1: MOV 26A ECCS 1:RHRSW C, pass -1 Hour 2-2 Downstream d MOV RB227D Mf d J6 L AN 26A FyroturrRLrarSIAFP
t ~ l Degradation Mechanisms Summary i i i i 10 welds with thermal fatigue j l I
- 2 welds with corrosion cracking l
1 I i 14 welds in IGSCC Program l l i 16 welds in FAC Program i i f f mourmun-
18 ul-96 g, gggg} pagjpg Hermal Fatigue ID: T-CS-01 A Applicable: Yes Comments: Thermal Fatigue is possible because welds are susceptilbe to Convetion Heating. Systems. CS 10-W23-l 504-5 A Welds. 10-14-471,10-14-472,10-14-473,10-14-473 A.10-14-474, N 5A SE hermal Tranden't No is the location a thermal sleeve? Yes There is a thermal sleeve at the injection point. Step 141: Is there cold water injection into a hot component? No Step l A: Is the material carbon steel or low alloy steel? No Step 1-B: Is the operating temp greater than 220F7 No Step l-C: Can the diff temp cxceed 150F7 No Step I.D: Is the material austenitic steel? No Step 1.B: Is the operating temp greater than 270F7 No Step 1-F: Can the diff temp exceed 200F7 No Step l-G: Does the expected detta T exceed the allowable ? No (TASCS) Low Flow / Valve Leakng No Step 2: Is the operating temp greater than 220F7 No Operating temperature is 150F or below. Step 2 A: Is the nominal pipe size larger than 1 inch? Yes Pipe size = 10" Step 2.B: Is the pipe slope < 45 degrees or elbow? Yes The pipe is sloped within 45 degrees of horizontal. Step 2-C: Is the (abs) value of the max temp diff > 50 deg? No Step 2-D: Is the max value of R1 > = 4.07 No ITASCS) Convection Heatinn Yes Step 3-A: Is the piping dead-ended? Yes Pipe is dead <nded due to the closed check valve AOV. 13A. Step 3-F: Is the nominal pipe size > 1 inch? Yes Pipe size = 10* Step 3-B: Is the pipe slope within 45 deg of horizontal? Yes The pipe is sloped within 45 degrees of horizontal. Step 3-C: Is a heat source available? Yes RCS is available as a heat source. Step 3-D: Is convective heat transfer possible? Yes There are no closed valves between this location and the heat source. Step 3-E: Is a temp diff of 50 deg likely between locntion and Yes nere may not be a temperature difference of 50 degrees heat source? between the closer locations and the heat souce (Reactor Vessel). ney have, however, been included since the entire horizontal section is assumed to have a potential temperature difference of greater than 50 degrees from reactor vessel temperature. (TASCSISteamAYatgr No Step 4; 1s the nominal pipe size > 1 inch? Yes Pipe size = 10" 3 4 Step 4-A: Is the piping connected to a component with steam? No j Step 4 B: Could steam propagate to the location? No Are Fatigue Analyses applicabic? No
4 18-Jul-96 DM - Corrosion Cracking Corrosion Cracking ID: C-CS-01 Applicable: Yes Comments: Corrosion Cracking is susceptialbe to Crevice Cracking. 4 Systems. CS lines: 10-W23 1504-5A //10-W231504-5B Welds. i N-5A SE// N-5B-SE Chloride Cracktne (Internath No is a water chemistry program in piace? Yes See Section 8 of the JAF Chemistry Manual for details. Step l A: Is the material'austenitic stainless steel? No 2 l Step l B: Is the operating temp greater than 120F7 No Step l C: Is the process fluid brackish or untreated water? No i i Step 1 D: Is inleakage from connected systems with No contaminants possible? Step l E: Is there low flow, intermittent flow or stagnation? No Step l-F: Is there a history / potential for comtamination by No < chlorides, fluorides, sulfides, etc. j Crevice Corrosion: Yes Step 2-A: Is the location a thennal sleeve? Yes The location is a reactor vesselinlet nozzle. These nozzles are fitted with thermal sleeves (ref. FSAR pg. 4.2-3). Step 2-B: Is the process fluid liquid water? Yes Reactorcoolant Step 2-C: Is the pressure boundary material steel? Yes Pipe Class 1504 SS Step 2 D: Is the pressure boundary material a nickel based No alloy? j Chloride Crackine &xternalk No Step 3 B: Is the material austenitic stainless steel? No Step 3 A: Is the piping insulation not in compliance with RG No JAF has a program in place to ensure that all piping 1.367 insulation located on austenitic stainless steel piping mets RO !.36. Step 3-C: Is the location within SD of a probable leak path? No Step 3-D: Is the piping exposed to wetting from a chloride No All piping is located either in the Reactor Building or the bearing environment such as salt water mist for coastal plants? Drywell. 2I 2 p I
Degrsdation Mechanism Consequence Risk No. of Segment Elements Type Category Description Category Cetegory Region CS-01 2 IGSCC,TF Small Leak LOCAin Drywell Medium Category 5 Medium -IGSCC Program CS-13 &C CS-02 4 IGSCC & Small Leak LOCAin Drywell Medium Category 5 Medium -IGSCC Program CS-14 TF CS-03 2 TF Smail leak LOCAin Drywell Medium Category 5 Medium CS-15 s CS-04 2 IGSCC & Small Leak LOCAin Drywell Medium Category 5 Medium - IGSC C Program TF El CS-05 7 IGSCC Small Leak LOCAin Drywell Medium Category 5 Medium CS-16 CS-06 13 None LOCA!n Drywell Medium Category 6 Low ': CS-17 CS-07 9 None Loss of CSTrain A Low Category 7 Low-CS-19 CS-08 4 None LOCAOutside of High Category 4 Medium CS-19 Drywell,1 Isolation Valve CS-09 136 None lcolable Loss of Low Category 7 Low. CS-20 Suppresabn Pool with Possible Flood to Both Crescent Areas ^ CS-10 20 None Loss of Suppression Medium Category 6 Low 2 CS-21 Poolin Torus Room CS-11 3 None isolable Loss of Low Category 7 . Low' CS-22 Suppression Pool with Possible Flood to Both j-Crescent Areas 4 CS-12 16 FAC Large Leak Loss of Suppression Low Category 5 Medium - FAC Program CS-23 Pool Pressure Boundary ? . ~
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> s, . =ro r0 m> j [ ;.:.p..:.: u A.m n. ?0 gi r .mo* w.oey>0 a ~ .:.:.:.:!.!I!j 3 i e O j .q X-n ao> a h. 5g ve > . emo".oh1sg A*.e n 7:::..i:::.:: :. 1:. .s .ii j iit... -~ a TI$ 5 h g & a 5-u t:!I ll1l 1l l l' [,
(ESW-C-01 A) PUMP RM A & SUCTION SW260E MOV-[102 A BAY (ESW-C-06A) t ESW P-2A EDG2728 { EDG PIPE CHASE STRAINER kj ESW LOADS 1f MOV-;101 A (ESW-C-07A) n EAST CABLE TUN 5EL f .... 7... y.: (ESW-C-02A) d.. { { ESW-C-05 A) SW260E EAST CABLE TUNNEL [EDG272A B C ~ X EDG X 4A SA DISCHARGE (ESW-C-03 ) TUNNEL EDG PIPE CHASE X X I V2A ....4 t A ....V A2B (ESW-C-04 ) EDG N EDG 272c 48 5B DISCHARGE TUNNEL l (ESW-C-02B) !(ESW-C-058) 4D SD
- EDG272B 8 D
~ EAST & WEST CABLE TUNNEL DG y PE R R.. _--_--......-A-..,i-ESW P-2B M (ESW-C-07B) EAST CABi.E TUCNEL i \\ p ESW LOADS sW260E STRAINER 1B MOV,101 B -(ESW-C-06B) O EDG PIPE CHASE PUMP RM B (ESW-C-01 B) - SUCTION SW2600 BAY MOV.1028 EMERGENCY SERVICE WATER FIGURE 5-17
30-Sep.96 FMECA - Consequence Inforniation Report Consequence ID: ESW.C-03 Pipe Section Descrintion: EDG A & C Supply in EDG Pipe Chase i f Spatial Affects: Propagation j Affected Location: EDGPC Spatial Effects Comments: EDG pipe chase fills and propagates into EDG B room (EDG272B). Eventually water will spill into adjacent electrical room and fail EDG B & D. l .I Initiating Event: No j Initiating Event Isolability: PotentialInitiating Event: PotentialIE Prevention: Configuration: Test or Demand (Periodic Testing). Demand evaluated as worst case (LOSP assumed). l l DirectImpact: EDG A & C (flow diversion) which fails train A systems. l Indirect Impact: EDG B & D if not isolated. Avall. Means ofIsolation: Trip pump before EDG B & D fail. There is no inunediate detection capability in this area. Avall. Trains of HP Makeup: RCIC + HPCI + SRVs = 2.5 trains Avail. Trains of LP Makeup: None (no credit for using diesel fire water) j kvailTrains of Heat Removal: CV = I trains Containment Performance: Unaffected l Other Operating Modes: Similar consequences during shutdown, but notjudged to be more significant. External Events: For seismic, one can assume the seismic capacity of pipe is high and the freqeuncy of challenge is comparable or less than the frequency used in this evaluation. Similarly for fires, the frequency of challenge should be comparable or less than the frequency used in this evaluation. I Consequence Category: HIGli Consequence Comments: Demand impact is "HIGH" in Table 2 2 based on unexpected frequency of challenge, "between test" exposure time, and no backup train (isolation before EDG B & D flooded not credited). 1 1 28 1
O 30-Sep-96 FMECA - Consequence Information Report Consequence ID: ESW-C-04 Pipe Section
Description:
EDG B & D Supp!y in EDG C Room Spatial Affects: Propagation Affected Location: EDG272C l Spatial Effects Comments: Eventually water will spill into adjacent electrical room and fail EDG A & C. Initiating Event: No Initiating Event Isolability: PotentialInitiating Event: f PotentialIE Prevention: Configuration: Test or Demand (Periodic Testing). Demand evaluated as worst case (LOSP assumed), i Direct Impact: EDG B & D (flow diversion) which fails train B systems. Indirect Impact: EDG A & C if not isolated. Avail. hicans of Isolation: Trip pump before EDG A & C fail 'Ihere is no know immediate detection capability in this area. Avall. Trains of HP hiakeup: RCIC + HPCI + SRVs = 2.5 trains Avail. Trains of LP hlakeup: None (no credit for using diesel fire water) tvall Trains of Heat Removal: CV = 1 trains Containment Perfonnance: Unaffected 'Other Operating hIodes: Similar consequences during shutdown, but notjudged to be more significant. External Events: For seismic, one can assume the seismic capacity of pipe is high and the freqeuncy of challenge is comparable or less than the frequency used in this evaluation. Similarly for fires, the frequency of challenge should be comparable or less than the frequency used in this evaluation. Consequence Category: HIGH Consequence Comments: Demand impact is "HIGH" in Table 2-2 based on unexpected frequency of challenge, "between test" exposure time, and no backup train (isolation before EDG A & C flooded not credited). l i l 29 l
P:gs 81 Calculation No, NSD-017 Table 5-16 Emergency Service Water Pipe onsequence ID Seedon Spatial Initisting Lost Available Containment Exposure Table Description Location Configuration Event (IE) Isolation Trains Traias EITeets Time Used Rank ESW-C-01 A Emergency Service SW260E Demand Anw id na AC Train A W AC Tram Unacected Bdmen 2-2 MEDIUM LOSP B Tests Water Pump 2A Test loss of 1: Trip Pump
- None, All, but tram Unaffected Test &
2-3 LOW PCS normal AC A depends on AOT available normal SW & AC ESW-C-OlB Emergency Service SW260D Dsne,d Assumed na AC Train B W ACTrain UnatYected Between 2-2 MEDIUM LOSP A Tests Water Pump 2B Test Loss of 1: Trip Pump
- None, Ali, but train Unattected Test &
2-3 LOW ? PCS normal AC B depends on AOT available normal SW & AC ESW-C-02A EDO A & C Supply in ECT256 Ds 4eid Assumed na AC Train A W AC Train Unailected Bd-cca 2-2 MEDIUM East Cable Tunnel LOSP B Tests ESW-C-02B EDG B & D Supply in ECF256 Demand Assumed na AC Train B W AC Tram Unadected Bdmse 2-2 MEDIUM East & WestCable WCT256 IDSP A Tests f l Turmels ESW-C-03 EDO A & C Supply m EDGPC Dweid Assumed Trip Pump All AC None Unaticcted Between 2-2 li!GH L Tests LOSP EDG Pipe Chase ESW-C-04 EDG B & D Supply in EDU272C Demand Assumed Trip Pump A11AC None Unatiec:ed Between 2-2 111G11 Tests EDG C Room LOSP ESW C-05A EDG A & C Supply & EDG272A Demand Anm.ed na AC Train A W AC Train UnafIected Between 2-2 MEDIUM Discharge EDG272C LOSP B Tests ESW-C-05B EDO B & D Supply & EDG272B Demand Assumed na AC Train B W AC Tram Unaffected Between 2-2 MEDIUM Discharge EDG272D LOSP A Tests r EDGPC ESW C-06A EDG A & C '.scharge EDG272B Demmd Assumed Trip Pump All AC None Unaffected Between 2-2 il1Gil Tests in EDG B Room & EDGPC LOSP EDG Pipe Chase ESW-C-06B EDG B & D Discharge EDOPL Demand Assumed na AC Train B W AC1 rain Unatfected Between 2-2 MEDIUM LOSP A Tests ESW-C-07A EDO A & C Discharge ECT256 Demand Assumed na AC Train A W AC Tram Unaficcted Between 2-2 MEDIUM i in EDG Pipe Chase in East Cable Tunnel SW260F LOSP B Tests ESW-C-07B EDO B & D Discharge ECT256 Demand Assumed na AC 1 rain B W AC Tram Unaffected Between 2-2 MEDIUM & Screenwell in East Cable Tunnel SW260E LOSP A Tests & Screenbli f I Draf19/30/96 [
Calculation No.NSD-0?? Figure 5-3 A Emergency Feedwater (EFW) Suction s s EFW-C-11 Room 2025 l ,~---------------] p, -816 Outside s i Room 2055 i 2EFW-2B r SW CST i 2C, -345 l f' i Room 2040 h 2CV 0776-1 HDRL i T41B 2CS-344~ 2 ________a,,,,,_,,,,,,,,,,,,,,,,,,,_,i e EFW-C-08B 2CS-817 'a e e e Room 2025 i i e e ppw.c ogn j l 2EFW-23 Room 2025 Stanup & B/ D Demin -{><Q-- To APN l ms 7' 2EFW-0706 I 2CT-5 X X N N N w 2Cv-0789-i CST I 2T41A 2CV-0707 2P7B 2EFW-16 2EFW-801 2EFW-1 2EFW-8)2 i l l EFW-C-10 l 2CT-41 Outside l'-~~--------------- ---,-,,, g - g,'] CST Room 2223 l 2T41B i 2CT-40 Room 2225 l i i ' EFW-C-09A 2d\\,-0795-2 l Room 2050 ? i l Room 2024 i i i EFW-C-08A Room 2024 iP7A 8 g I g i i g 5 2E,FW-2A SW us /s r., HDR2 l 2CVq711-2 Draft 2/28/97
rageav Celculation No. NSD-0?? Figure 5-3B Emergency Feedwater (EFW) Pump Discharge 1 l 2EFW-4B N x 7 '2CV-1025-1 EFW-C-05B MDP [- 2EFW-6 Room 2040 2P7D FFW-C-06B Room 2081 i e Room 2025 .i 8 Room 2055 i1, i Room 2084 From AFW .Q q I V,2EFW-5A___________,,lf i 2CV-714,
- 2CV-1075-1 2EFW-) l _,, _ _ _ _ _ _ _ _ _ _ _ _ _ _
JL_______._._-______, 2EFW-10B' i l l FFW-C-07
- i
._._V V ~. Room 2025, _. 2 L. _ _JL. _ _.i Room 2024 e 2EFW-29_ _ _ _l~_ _ _ _ :-- ::_ _ _V-
- _ _ :: _l 2ggy.g g3 l
N J k 2EFW-5B From AFW, l k q l. EFW-C-05A CV-1026-2 2EFW-4A Room 2040 2EFW-OA+- 7-l. Room 20SI EFW-C-06A l' Room 2055 k Room 2024 l Room 2084 i 7' ...._-._.....--__.-_.-___--..-____..e i .----___.....___.-_...._____i 2CV-1076-2 Draft 2/28/97 4 i
b Page 51 Calculation No. NSD-0?? Figure 5-3C Emergency Feedwater (EFW) Discharge to Main Feedwater & SGs Inside Containment 1FW C-04E EFW-C-03B Room 2084 g" g Room 2084 2M'-9A X ,Ny ]l ,NQ To SO "A" 2CV- 0251 2C%1038-2 2EF V-7B EFW.C-02A N.I2A Room 2C34 A EFW-C-03A m 208 EFW-C-Old ) Room 208( & 2 r, s s s 2CV-1026-2 2C% 1037 1 2EFW-7A l 2P7B --> l EFW-C-04E EFW-C-03D 2P7A - ---> Room 2081 a Room 2081 7EFW-9B To SG "B" uY2 s s% s r, ys 8 s 2CW 075-1 2CV-1036-2 2E V-8B J EFW.C-02B EFW-C-12B Room 2C81 l PW-C-04C l l oom 208 EEEC-03C EFW-C-01B D Room 208( i l m,
- r.,
8 s N 2CW 076-2 2CW1039-1 2EFW-8A Inside Containment d l
aw Calculation NO.NSD-0?? Tabic 5-3 Consequence Analysis Summary -Emergency feedwater iD Description Spatial Configuration initiator isolation Systern impacts Backup Trains Containment E-T- - 4 Table Rank Location (note 1) (note 1) Tinw Used EF W-C-01 A EFW to SG A Contamment Demand Assumed 2EFW-9A PCS O2), EFW 2(EFW A AFW 2EFW-7A A kimen 2-2 MEDIUM Inside T2 Inside & to SG "A" to SG "B", Once 7B outside test Containment MOVs n Through m outside PCS, EFW, isolation & Once AFW Through) Demand Assumed 2EFW-9B PCS G2) EFW 2 (EFW & AFW 2EFW-8A & between 2-2 MEDIUM EF W-C-01 B EFW to SG B Containment T2 inside & to SG "B" to SG "A", Once SB outside test Inside Containment MOVs m Through m outside PCS, EFW, Isolation & Once AFW Through) EF W 4'-02A EFW to SG A 2084 Demand Assumed 2EFW-9A PCS (T2), EFW 2(EFW & AFW 2EFW 9A kiw en 2-2 MEDIUM Outside T2 Inside & to SG "A" to SG "B", Once inside test Containment MOVs m Through s outside PCS,EFW, Isolation & Once AFW, Once Through B) Through A EF N-C-028 EFW to SG B 2081 Demand Assumed 2EFW-98 PCS G2), EFW 2 (EFW & AFW 2EFW-9B between 2-2 MEDIUM Outside T2 inside & to SG "B" to SG "A", Once inside test C%inment MOVs a 1hrough m outside PCS, EFW, Isolation & Once AFW, Once Through B) W t Through A EF W-C-03 A EFW to SG A 2084 Demand Assumed 2CV-1026 PCS G2), EFW 3 (EFW, AFW, Unaffected between 2-2 LOW between 2CV-T2 or 1037 B to SG "A" Once Through a test l 1037 and 2EFW. m isolation, EFW B PCS,EFW A. & Once Through 7A AFW. Once B) Through A EF W-C-03 B EFW to SG A 2084 Demand Assumed 2CV-1025 PCS G2), EFW 2.5 (EFW, AFW, Unaffected between 2-2 LOW between 2CV-T2 or 1038 A to SG "A" Once Through m test 103S and 2EFW-g isolation.EFW PCS, EFW B. A & Once 7B AFW, Once Through B) Through A EF W-C-03C EFW to SG B 2081 Demand Assumed 2CV-1076 PCS G2), EFW 3 (EFW, AFW, UnatTected between 2-2 LOW between 2CV-T2 or 1039 A to SG *B" Once Thrcugh m test 1039 and 2EFW-m isolation, EFW B PCS, EFW A. & Once Through 8A AFW, Once B) 1hrough A l ? Draft 2/28/97 --e e
aupw. Calculation NO.NSD-0?? Table 5-3 Cmsquen::e Analysis Summary-Emergency Feedwater ID Description Spatial Configuration Initiator isolation Systemimpacts Backup Trams Containment Exposure Table Rant Location (note 1) (note 1) Time Used EFW-C-03D EFW to SG B 2081 Demand Assumed 2CV-1075 PCS (T2). EFW 2.5 (EFW. AFW, UnatYected between 2-2 LOW between 2CV-T2 or 1036 B to SG "B" Once Through ct test 1036 end 2EFW. ct isolation, EFW BB PCS, EFW B, A & Once AFW, Once Through B) 1hrough A EF W-C-04A EFW to SG A 2084 Demand Assumed 2CV-1026 PCS (I2), EFW 3 (EFW, AFW, Unaficcted between 2-2 LOW T2 A to SG "A" Once Through cr test i between 2CV. i 1026 and 1037 ct isolation EFWB PCS,EFW A. & Once Through AFW,Once B) 1hrough A Ef W-C-04B EFW to SG A 2084 Demand Assumed 2CV-1025 PCS (I2), EFW 2.5 (EFW, AFW, Unairected between 2-2 LOW between 2CV-T2 B to SG"A" Oncelhrough ct test 1025 and 1038 ct isolation EFW PCS, EFW B, A & Once AFW,Once Through B) Through A EFW-C-04C EFW to SG B 2081 Demand Assumed 2CV-1076 PCS (F2), EFW 3 (EFW, AFW, Unaffected ber-c a 2-2 LOW between 2CV-T2 A to SG "B" Once Through cr test 1076 and 1039 ct isolation,EFW B PCS, EFW A. & Once Through AFW Once B) i Through A EFW to SG B 2081 Demand Assumed 2CV-1075 PCS(I2) EFW 2.5 (EFW. AFW, UnafIected between 22 LOW lI EFW{-04D between 2CV-I T2 B to SG "B" Once Through ct test p 1075 and 1036 ct isolation, EFW PCS, EFW B. A & Once AFW, Once Through B) Through A EF W-C-05A EFW A to SGs 2084~ Demand Assumed 2P7A trip PCS (I2), EFW 2.5 (EFW B, Unaffected Mr-cin 2-2 LOW Outside Pump 208? T2 & suction A.ct AFW, Once test Rooms 2N' MOVs PCS,EFW A. Through ct 2040 AFW, Once Isolation, EFW B Through A & Once Through j B) o EF W-C458 EFW B to SGs 2084 Demand Assumed 2P7B trip PCS (F2), EFW 2(EFW A, Unaffected ber.cw 2-2 MEDIUM Outside Pump 208i T2 & suction B ct AFW,Once test Rooms 2055 MOVs PCS,EFW B. Through ct 2040 AFW,Once isolation,EFW l, Through A A & Once Through B) I. i l Draf12/28/97 1
i Calculation NO. NSD-0?? Tsole 54 Consequence Analygis Summary. Emergency Feedwater ID Description Spatial Configuration Initiator Isolation Sp-Impacts Backup Trains Containment Exposure Table Rank Imcation (note 1) (note I) Time Used EF W-C-06A EFW A in Pump 2024 Demand Assumed 2P7A trip PCS U2), EFW 2.5 (EFW B. Unaffected ki.cca 2-2 LOW Room T2 or room A AFW,Once test flooding Through) EF W-C-068 EFW B in Pump 2025 Demand Assumed 2P7B trip PCS U2), EFW 2 (EFW A. Unaffected ks cca 2-2 MEDIUM Room 12 or room B AFW,Once test flooding Thmugh) EF W-C-07 Berween 2EFW-2024 Demand Assumed Yes (pipe PCS (T2), EFW 1(Once Unaffected altyear 2-2 LOW 5A and 5B 2025 T2 and challenge due to ehdienge Through) additiond requirs; oferosstics failures locally I (unexpecte opening d valves) ) frequency cf cha!!enge) EF W-C-08A EFW A Suction in 2024 Demand Assumed 2CV-0795 PCS (12), EFW 2.5 (EFW B. Unaffected between 2-2 LOW Pump Room T2 and 0711 A AFW,Once test and SWS Through) EF W-C-08B EFW B Suction in 2025 Demand Assumed 2CV-0789 PCS (T2), EFW 2 (EFW A. Unaffected between 2-2 MEDIUM Pump Room T2 and 0716 B AFW,Once test and SWS Through) EF W-C-09A Common CST 2024 Demand Assumed 2CV-0707 PCS U2), EFW 2.5 (EFW B. Unaffected between 2-2 LOW Pipe in EFW A T2 locallyand A AFW, Once test Pump Room SWS Through) EF W-C-098 Common CST 2025 Demand Assumed 2CV-0707 PCS(TT EFW 2(EFW A, Unaffected between 2-2 MEDIUM l Pipe in EFW B T2 locally and U AFW,Once test Pump Room SWS nrough) EF W-C-10 Common CST Outside Demand Assumed NA PCS (T2) 3 (EFW, AFW, Unaffected ki-cca 2-2 LOW Pipe Outside 2223 T2 Once Through) test l 2225 2050 EF W-C-11 Unit 1 CST Piping Oatside Demand Assumed NA PCS U2) 2 (EFW A. Unaffected all year 2-2 LOW upstrer.m of check 2055 T2 & Unit AFW,Once valves 844 A 845 2040 1 CST Through) 2025 challenge (unexpecte d f frequency of j challenge) EF W-C-12A EFW to SG 'A" Containment Standby TS No PCS (T5)& NA Unaffected NA 2-1 MEDIUM i downstream of EFW to SG "A" 2EFW-9A due to T5 l Draft 2/28/97 t i t
CalculationNo.NSD-0?? I i lable 5-3 Consequence Analysis Summary-Emergency Feedwater ID Description Spatial Configuranon initiator Isolation System impacts Backup Trains Containment Exyw-c Table Rank IAcation (note 1) (note 1) Time Used EF W-C.128 EFW to SG "B" Containment Standby T5 No PCS (TS) & NA Unaffected NA 21 MEDIUM downstream of EFW to SG"B" 2EFW-9B due tots Note 1: Successful isolation result is shown first and then isolation failure case separated by "of' when applicable. "Once Through B" refers to once through cooling mode ofinventory control and heat removal with train B discharge valves available (both pump trains are available but the A train discharge valves are failed). Similarly,"Once Through A" refers to train A discharge valves. i Draft 2/28/97 m b
-~ --. i 28-Feb-97 FMECA - Consequence Information Report C:nsequence ID: EFW-C-01A C:nsequence
Description:
Degradation of EFW flow to steam generator 2E-24A inside containment during an independent demand (line 2DDB-3 between containment penetration and check valve 2EFW-9A) Break Size: Large Isolability of Break: Yes ISO Comments: 2CV-1025-1 or 1038-2 and 2CV-1026-2 or 1037-1 can be closed by the operator to prevent dumping part of the CST into containment. Detection is based on flow indication from EFW pumps, with a continued lowering of steam generator level (i.e., faulted steam generator). Each of the 4 EFW discharge lines has flow indication and annunciation (>325 gpm and <240 gpm). Eventually, containment sump level would provide indication oflost steam generator inventory l due to EFW water not reaching the steam generator. SpatialEffects: Containment Effected Location: Containment Building Spatial Effects Comments: None. Equipment located within the containment is qualified. Initiating Event: N Initiating EventID: N/A Initiating Event Recovery: Pipe break during normal operation is unlikely since this piping is isolated from steam generator by 2EFW-9A. A loss of PCS (T2) initiator is assumed to challenge this piping. - Loss of System: SDM-2 System IPE ID: PCS, EFW System Recovery: PCS loss is assumed to be the initiator. Pipe degradation causes loss of EFW flow to steam. generator "A". Isolation failure is assumed to cause flow diversion from both EFW trains. Loss of Train: N Train ID: N/A Train Recovery: N/A Consequence Comment: Consequence is " Low" based on Table 2-2 (anticipated frequency of challenge, between test exposure time,3 backup trains - EFW, AFW, and once through cooling). For the i isolation failure case, the consequence is " Medium" with 2 backup trains (isolation failure and once through cooling). 2EFW-7A & 7B provide containment isolation outside the containment building, thus, the consequence remahis unchanged. j Consequence Category: MEDIUM O Consequence nank: 1 1 31 L
. --. ~. e i i 1 i 28-Feb-97 FMECA - Consequence Information Report Consequence ID: EFW-C-06B Cousequence
Description:
Degradation of EFW Pump 2P7B flow to both steam generators during an independent demand in Room 2025 (line 2DBC-1 and 2DBC-12 from AFW) Break Size: Large Isolability of Break: Yes ISO Comments: Pump 2P7B can be tripped oc flooding the room will trip the pump. There is flood detection in room 2025 with alarm in the control room. Propagation through the floor drain to therauxiliary building sump and its high level alarm provides additional detection capability. Failure of the pump due to room flooding and a watertight room provides the equivalence ofisolation. Spatial Effects: Propagation Effected Location: Room 2025 Spatial Effects Comrnents: The motor driven pump is assumed flooded and unavailable. Propagation is through floor drain to El 317 and any leakage out of the room is assumed to be within floor drainge capability outside. Initiating Event: N Initiating Event ID: N/A Initiating Event Recovery: I ipe break dudag normal operation is unlikely since this piping is isolated from the l tteam generators. A loss of PCS (T2) initiator is assumed to challenge this piping, s Loss of System: S System IPE ID: PCS j System Recovery: PCS loss is nssumed to be the initiator, Loss of Train: T Train ID: EFW "B" Train Recovery: Pipe degradation causes loss of EFW pump 2P7B flow to both steam generators. 1 Consequence Comment: Consequence is " Medium" based on Table 2-2 (anticipated frequency of challenge, j between test exposure time,2 backup trains - EFW "A", AFW, and once through cooling). No impact on containment isolation. Consequence Category: MEDIUM Consequence Rank: O l i d 4 46 1
o \\ 28-Feb-97 FMECA - Consequence Information Report Consequence ID: EFW-C-06A Cnsequence
Description:
Degradation of EFW Pump 2P7A flow to both steam generators during an independent demand in Room 2024 (line 2DBC-2 and 2DBC-13 from AFW) Break Size: Large Isolability of Break: Yes ISO Comments: Pump 2P7A can be tripped or flooding the room will trip the pump. There is flood detection in room 2024 with alarm in the control room. Propagation through the floor drain to the auxiliary building sump and its high level alarm provides additional detection capability. Failure of the pump due to room flooding and a watertight room provides the equivalence of isolation. Spatial Effects: Propagation Effected Location: Room 2024 Spatial Effects Comments: The turbine driven pump is assumed floolled and unavailable. Propagation is through floor drain to El 317 and any leakage out of the room is assumed to be within floor drainage capability outside. Initiating Event: N Initiating Event ID: N/A Iultiating Event Recovery: Pipe break during normt.1 operation is unlikely since this piping is isolated froni the steam generators. A loss of PCS (T2) initiator is assumed to challenge this piping. Loss of System: S System IPE ID: PCS System Recovery: PCS loss is assumed to be the initiator. Loss of Train: T Train ID: EFW "A" Train Recovery: Pipe degradation causes loss of EFW pump 2P7A flow to both steam generators. C:nsequence Comment: Consequence is " Low" based on Table 2-2 (anticipated frequency of challenge, between test exposure time,2.5 backup trains - EFW "B", AFW, and once through cooling). No impact on containment isolation. C nsequenceCategory: LOW D Consequence Rank: 0 \\, l 45 i
r
- n r.
L I i i i i 28-Feb-97 FMECA - Consequence Inforniation Report [ Consequence ID: EFW-C-12A t Ccusequence
Description:
Degradation of main feedwater flow to steam generator 2E-24A inside containment during normal operation (EFW line 2DBB-3 downstream of 2EFW-9A). j Break Size: Large Isolability of Break: No l ISO Comments: Feedwater isolation and feedwater pump trip will occur on low steam generator pressure. Also, j EFW will remain isolated to the faulted steam generator via a differential pressure between the j. faulted and good steam generators. However, blowdown of the faulted steam generator can not j be isolated. f Spatial Effects: Containment Effected Location: Containment Building )' Spatial Effects Comments: Feedwater line breaks are within the design basis and the necessary safety I {j components located inside the containment building are qualified for such events. { Initisting Event: I Initiating Event ID: T5 1 Initiating Event Recovery: No recovery from an unisolable feedwater line break. This results in an immediate plant trip due to low steam generator level. Loss of System: SDM 2 System IPEID: PCS, EFW f System Recovery: MSIV isolation, feedwater isolation and pump trips occur o'n low steam generator pressure. it is possible to recover a condensate pump and makeup to the unfaulted steam generator. EFW discharge to the faulted steam generator is isolated and unavailable. However, there is a discharge path from each EFW pump to the unfaulted steam generator. Loss of Trals: N Train ID: N/A Train Recovery: N/A Consequence Comment: Consequence is " Medium" based on Table 2-1. Containment isolation is unatTected. Consequence Category: MEDIUM O Consequence Rank: O f o> I 54
l t EPRI l l i f i r f > I PIPE FAILURE POTENTIAL via 1l1 z\\ DEGRADATION MECHANISM ASSESSMENT ?\\ Ht i n! i S. R. Gosselin, P.E. Electric Power Research Institute t i Component Integrity Technology see SOO42%7. PPT - 15 t
1 i b EPRI - 7 i I i insights from Service Experience The likelihood of service induced pipe failures (cracks, leaks, or i breaks) is strongly dependent upon the presence of an active j degradation mechanisms. Examples include-l Thermal Fatigue (TF) l Corrosion Fatigue (CF) l Stress Corrosion Cracking (SCC) - e.g., IGSCC, PWSCC, etc. Erosion-Cavitation (E-C) l
- Corrosion Attack (COR) - e.g., MIC, Crevice Corrosion, Pitting, etc.
Erosion - Corrosion (E/C) - i.e., FlowAccelerated Corrosion High Cycle Mechanical Vibration Fatigue (VF) I Failures do not generally correlate with stress or fatigue usage factot values contained in Design Reports Failures do not always occur in welds i 1 i Component Integrity Technology SG042997. PPT - 16 l
+ - EPRI Insights from Service Experience l Case Study ASME Class 1 Category BJ Welds i Survey involved 50 plants 733 cumulative years of operation l 37,332 Category BJ welds 3 Task Group Conclusions Total of 156 Category BJ welds were found to contain service induced Daws l 151 - IGSCC (currently addressed by NRC Generic Letter 88-01) i 3 -Fatigue 2 - General Corrosion 55 failures were detected by ASME Section XI examinations (101 failures were detected by augmented examinations) l Assuming a totalISI population of 9333 (25% of total) l 0.59% of the welds examined by ASME Section XI contained flaws 0.05% of the welds inspected contained flaws caused by a mechanism other than IGSCC l None of the flawed welds fell into the category of "high stress" of"high fatigue usage" l i i Coniponent Integrity Technology s I SG042997. PPT - 17 l [
l - EPRI l l Identifying Degradation Mechanisms Degradation mechanisms are identified by explicit. well defined atitihutes which provide the means for a uniform. consistent evaluation Generic attributes have been determined for each degradation mechanism based on industry experience, basic casual effects, EPRI studies, and NRC requirements j Generic effects are used on a pipe system / segment specific basis to identify the presence of potentially active degradation mechanisms and and their locations i i Component Integrity Technology s_. SG042997. PPT - 18 i
- EPRI Service Failure Data l SKI Report 96-20 " Piping Failures in United States Nuclear Power [ Plants: 1961-1995", January 1996 l r 1511 Piping and Pipe Component Failure Events j - Encompassing 2068 reactor operating years { Failure Mechanism Number of Failures l l 1 Corrosion Fatigue (CF) 14 . Thermal Fatigue (TF) 38 Stress Corrosion Cracking (SCC) 166 l Erosion-Cavitation (E-C) 15 j Corrosion (COR) 72 l Design & Construction (D&C) 192 [ Water Hammer (WH) 35 Vibration Fatigue (VF) 364 Erosion / Corrosion (E/C) 280 l Other (OTH) 43 i Unknown 177* l
- Piping failures reported with unknown causes l
ComponentintegrityTechnology l h SG042997. PPT - 19 i
[ - EPRI Service Failure Data i FAK CATEGORY Cracks / Leaks - Flaws that penetrate the pipe wall and result in l visible signs of leakage (i.e., boric acid buildup, seepage, l dripping, etc.) Leaks - Wall penetration, finite leakage ranging from cc/hr to <1 l l 9Pm. Fallures - Generally larger leaks at rates greater than allowable l Technical Specification limits,1-10 gpm RUPTURE CATEGORY i Rupture, Breakage, & Severed-Pipe breaks occurred in a j significant portion of the pipe cross-section. In many cases, j actual leak rate was not recorded. i i i Component Integrity Technology s SG042997. PPT - 20 i
EPRI Failure and Rupture Frequency Estimates l Failure Frequency Conditional Rupture Failure Iper reactor Rupture Frequency Mechanism Failures Ruptures yearl Probability fper reactor yearl A {F} P{RlFf A{R} nj{F} nj{R} i j j Point Point Estimate Point Bayes Update Estimate Estimate Mean Value 4 CF 14 0 6.8 x 10 <0.071 * <4.8 x 10* 3.8 x 10'5 l TF 38 0 1.8 x 10'* <0.026* <4.8 x 10'4* 3.8 x 10-5 t SC 166 0 8.0 x 10-2 <0.0060* <4.8 x 10* 3.8 x 10~5 4 E-C 15 0 7.3 x 10 <0.067* <4.8 x 10* 3.8 x 10-5 COR 72 3 3.5 x 10-2 0.042 1.5 x 10* 1.2 x 10 i f WH 35 15 1.7 x 10-2 0.43 7.3 :s iC 6.8 x 10 4 E/C 280 19 1.4 x 10 O.068 9.2 x 10* 8.7 x 104 VF 364 25 1.8 x 10 O.069 1.2 x 10-2 1.2 x 10 2 OTH 43 8 2.1 x 10-2 0.19 3.9 x 10 3.5 x 10* [ 4 D&C 192 13 9.3 x 10'* 0.068 6.3 x 10 5.9 x 10* 4 UNK 177 11 8.6 x 10-2 0.062 5.3 x 10 4.9 x 10 4 4 TOTALS 1396 95 6.8 x 10 O.068 4.6 x 10 2 4.5 x 10-2
- Point estimate is zero based on no occurrences of ruptures, upper bound estimated assuming 1 rupture for each failure mechanism.
Component Integrity Technology e-SG042997. PPT - 21 l
fi' mTE"" 1 M TO s* @ g,3Ng* E ge aQ$3k= 4O @M [-g n h = ,og3E3 N ?Rk l y cn?Rb i e l u l l q l l i e l l i r F l l l l i e &}I g l l l l i ru l tp l l l l l i u?Rb l l 0 l l l t R l l l l l l l l l l l i l l l l l i l l l l l t i l l l hl l l l l l I i l l l l l l l l l l i 4 l l l l l l l l t ?Rk L C F o c R C F H D K M D C T s O / V T C N E a D C O U kig: fg53 t
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E EPRI - i t Bayes Update Pipe Rupture Frequencies for Different i Failure Mechanisms i t t Y l 1.00E-01 l l f E AMENABLE TO ISi D NOT AMENABLETO ISI i 1.00E-02 5 m E E g 6 x E $1. g E-03 E g g g E E E g n m m E g 1.00E-04 m a a m m m m m fj 1.00E-05 E-C SC TF CF COR OTH UNK D&C WH E/C VF FAILURE MECHANISMS Component Integrity Technology e.= SG042997. PPT - 23 (
- EPRI Degradation Mechanism Rupture Failure Potential Categories Larne Pipe Leak Conditions Dearadation Mechanism f Break Potential HIGH Larae Erosion Corrosion (FAC) Vibration Fatiaue (VF) j Waterhammer (WH) i i Thermal Fatiaue Corrosion Fatiaue/Crackina i MEDIUM Small Stress Corrosion Crackina (IGSCC, TGSCO,PWSCC, ECSCC) Local Corrosion Attack (MIC, Og [ Pittina) Erosion / Cavitation LOW None No Dearadation Mechanisms i For piping segments with the potential for catastrophic failure, the consequence analysis should be l conducted consistent with the plant's design basis. The design basis should determine whether a large leak t or catastrophic failure is appropriate. Absent this information, alternatives include use of the IPE/PRA or other engineering basis for establishing break characteristics. i Component integrity Technology SG042997. PPT - 24 [ i i
EPRI i i i i i l i RISKIMPACT-A QUALITATIVE 3l EXAMINATION ?l u, t -t 4l S. R. Gosselin, P.E. l Electric Power Research Institute l 3 I i i 4 t Component Integrity Technology in _ SOO42997. PPT - 25 i
7 i i - EPRI i Current ASME Section XI inservice Inspection Requirements ? Class 1 Piping [ >1" NPS - Volmetric & Surface Exams on 25% welds each 120 month inspection interval. All piping - Leakage Tests (VT 2) each refueling outage. l Class 2 Piping >2"NPS ECCS Piping (>4" NPS Non-ECCS Piping) Volumetric and Surface Exams on 7.5% l of wieds each 120 month inspection interval j All piping - Leakage Tests (VT2) each refueling outage. Class 3 Piping All piping - Leakage Tests (VT2) each refueling outage } All piping - Hydrostatic Pressure Test each 120 month inspection interval. Non ASME Section lli Piping l - No Section XI inservice inspection requirements Component integrity Technology S0042997. PPT - 28
e s EPRI ) Alternative Risk-informed i Inservice inspection Requirements t t Class 1 Piping l >1" NPS - Replace existing requirements for Volumetric and Surface exams with alternative { RI-ISI criteria i All piping - Continue existing leakage tests and visual inspections each refueling outage. { Class 2 Piping j Replace existing requirements for Volumetric and Surface exams with alternative RI-ISI l criteria and expand scope to include all piping > 1" NPS l All piping - Continue existing leakage tests and visual examinations each refueling outage. l Class 3 Piping Expand existing requirements to perform Volumetric and Surface exams according to RI-ISI critena j All piping - Continue existing leakage tests and visual examinations each refueling outage. l All piping - Continue existing hydrostatic pressure test each 120 month inspectica inte mal. l Non ASME Section lli Piping Expand program scope to include Volumetric and Surface exams for non-code piping i (determined to be important to safety) according to alternative RI-ISI critena j i Component Integrity Technology l er SG042997. PPT - 27 l
-EPRI I 1 ANO SYSTEMS i b SYSTEM TITLE CODE CLASS 1 Reactor Coolant System RCS 1 2 Chemical Volume & Control CVCS 1, 2 l 3 High Pressure Safety injection HPSI 1, 2 4 Low Pressure Safety injection LPSI 1, 2 5 Shutdown Cooling SDC 1, 2 j 6 Containment Spray CS 2 7 Main Steam MS 2 8 Main Feedwater MF 2, 3, NNS 9 Emergency Feedwater EFW 2, 3, NNS 10 Service Water SW 2, 3, NNS i i t i 1 5 I I Component integrity Technology im S0042997. PPT - 28 i
r - EPRI ~ i ANO2 RCS l l (PreliminaryResults) t t i t NUMBER OF NUMBER OF CURRENT UT PROPOSED UT i SEGMENTS WELDS SXI RBI 1 i i HIGH 6 43 11 12 t l MEDIUM 26 229 38 23 LOW 8 35 2 0 TOTAL 40 307 53 35 l l l Component Integrity Technology n SOS 42997. PPT - 29 r .__j
i EPRI ANO2 RCS Risk Matrix l LCONSEQUENCE CATEGORY = t J; Core Melt Potential for Limiting Break Size) I I NONE LOW MEDIUM Birdi l t L /$R$218 !)) *..? Total = 0 f Totes eo : l f '"l l EO; i HIGH current = 0 current-o; .! e i jf! 3-!? New = 0 LNew = 0 : l L $ 1 3::(::g.: e + f
- I Total = 0 Total = 13 iTotal = 0 :
hlI l i l[!)l:i. MEDIUM current = 0 current = 2 currentioj [i, New = 0 New = 0 a..New = 0 / e .jp(::e.y-l r J O':R L A Y.:- lEE Total = 20 Total = 0 Total = 2 Total = 229.. ]{y#: LQE current = 0 current = 0 current = 0 current = as t (g: New = 0 New = 0 New = 0 ( New = 23 ; j =~ I i i Component Integrity Technology i m. SGou997. PPT - 30 l
i Mechanism Specific Examination Volumes and Methods t Acceptance Standard: Section XI, IWB 3514 Evaluation Standard (as applicable): Section XI, IWB - 3640 or IWB - 3650, ( Profilo of component x' W 4- \\ \\ 1/2 in. +/ J ',/ } g; \\ l Weld buttering f.. (where applicable) s_. T 1 t _{_ B j___ i l 4 MM7WM c f A' l Q 1/t Examination 3 Volume A-B-C-D Figure 7.1-1 Examination Volume for Thermal Fatigue Cracking in Piping Welds less than NPS 4. 1 l t l Profile of component l / / Weld buttering J,. ' (where applied) j { l 1 l \\ l 26 \\ \\ 222 { g B.
- 2
_C ~ 5 fvv0 -- - 2 1 t - -(s%q 4 ) A' o i 1/4" u e -B-C-D 1/4" Figure 7.12 Examination Volume for Thermal Fatigue Cracking in Piping Welds NPS 4 or Large l l 7-4 l =.;.;- ....m \\
l - EPRI J. A. Fitzpatrick Systems 4 SYSTEM TITLE CODE CL. ASS l 1 Reactor Water Recirculation RWRS 1 I 2 Main Steam MS 1 3 Main Feedwater MF 1 l 4 Core Spray CS 1, 2 l 5 Reator Water Cleanup RWCS 1 l 6 Control Rod Drive CRD 2 l 7 High Pressure Coolant injection HPCI 1, 2, NNS i 8 Residual Heat Removal RHR 1, 2 9 Reactor Core Isolation Cooling RCIC 1, 2, 3, NNS 10 Nuclear Boiler Vessel instrumentation INST 1 11 Standby Uquid Control SLC 1, 2 l 12 Fuel Pool Cooling FPC 3 13 Service Water & RHR Service Water RHRSW 3, NNS 14 Emeraency Service Water ESW 3 l Component Integrity Technology a SOM2997. PPT - 31 J
l EPRI J. A. Fitzpatrick System, Segments, and Welds j l Sys. No. System Safety Class No. of Locations No. of Segments 1 RWRS 1 142 60 2 MS 1 144 32 3 FW 1 81 13 4 CS 1,2 218 25 5 RWCU 1 36 7 [ 6 CRD 2 54 3 l 7 HPCI 1,2,NNS 212 27 8 RHR 1.2 887 1 01 9 RCIC 1,2,3,NNS 114 15 l 10 INST 1 25 6 i 11 SLC 1,2 21 4 l 1 12 FPC 3 30 7 l 13 RHRSW 3,NNS 37 10 [ 14 ESW 3 42 7 l l Component Integrity Technology a soum7.wr - 32
- EPRI l [ J. A. Fitzpatrick High Risk Region Result Summary i i i System No. of Locations Consequence Rank Deg. Mech. ID Risk Significance j RISK CATEGORY 1 } MS 44 HIGH FAC HIGH i FW 48 HIGH FAC HIGH l RWCU 2 HIGH FAC HIGH l RISK CATEGORY 2 HPCI 11 HIGH TF HIGH RHR 8 HIGH CAV,TF HIGH RCIC 8 HIGH CC HIGH RHRSW 11 HIGH MIC HIGH i ESW 4 HIGH MIC HIGH RISK CATEGORY 3 MS 74 MED FAC HIGH j FW 33 MED FAC HIGH RWCU 8 MED FAC HIGH HPCI 2 MED FAC HIGH 1 Component Integrity Technology m-soo42997. PPT - 33 i l l l
- EPRI ~ i J. A. Fitzpatrick Medium Risk Region Result Summary i i System No. of Locations Consequence Rank Deg. Mech. ID Risk Significance i RISK CATEGORY 4 [ MS 6 HIGH NONE MED } CS 8 HIGH NONE MED l RWCU 12 HIGH NONE MED l HPCI 11 HIGH NONE MED RHR 16 HIGH NONE MED RCIC 25 HIGH NONE MED l SLC 1 HIGH NONE MED RISK CATEGORY 5 i RWRS 112 MED IGSCC,CC .MED CS 17 MED IGSCC,TF MED HPCI 26 MED TF, CC, MIC MED R' R 17 MED-IGSCC, TF MED RU 3 6 MED MIC MED l IN67 1 MED TF MED SLC 2 MED TF MED l ESW 38 MED MIC MED MS 4 LOW FAC MED CS 16 LOW FAC MED i RWCU 13 LOW FAC MED RHR 74 LOW FAC MED i 1 Component Integrity Technology a SG042997. PPT - 34 F i
- EPRI ~ l J. A. Fitzpatrick Inspection Program Comparisons i i l No. of SXI RilSI Other Risk Category Locations Program Program Programs CAT 1, 2 & 3 253 2 11 211 i CAT 4 & 5 405 37 16 247 I CAT 6 & 7 1385 143 0 51 i Total 2043 182 27 509 Component Integrity Technology w SOS 42997. PPT - 35 t
- EPRI Qualitative Aspects of RI-ISI Inspection program defined according to pipe risk regardless of ASME Code Class Focus inspections in higher risk regions Implementing " inspection for cause" principles Inspection locations, examination methods, and acceptance criteria specific to degradation mechanisms known to be present Expanded inspection volumes Establishes an integrity management program based on an i understanding of piphg degradation mechanisms and their I relationship to the design, operation and maintenance of the plant i Component Integrity Technology a S0042997. PPT - 36}}