ML18151A577
| ML18151A577 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 08/31/1994 |
| From: | Doctor S, Gore B, Simonen F, Vo T Battelle Memorial Institute, PACIFIC NORTHWEST NATION |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-B-2289 NUREG-CR-6181, PNL-9020, NUDOCS 9409090115 | |
| Download: ML18151A577 (69) | |
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NUREG/CR-6181 PNL-9020 A Pilot Application of Risk-Based Methods to Establish In-Service Inspection Priorities for Nuclear Components at Surry Unit 1 Nuclear Power Station Prepared by T. Vo, B. Gore, F. Simonen, S. Doctor Pacific Northwest Laboratory Operated by Battelle Memorial Institute Prepared for U.S. Nuclear Regulatory Commission
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I 9409090115 94083!
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A Pilot Application of Risk-Based Methods to Establish In-Service Inspection Priorities for Nuclear Components at Surry Unit 1 Nuclear Power Station
~anuscript Completed: January 1994 w-'ate Published: August 1994 Prepared by T. Vo, B. Gore, F. Simonen, S. Doctor Pacific Northwest Laboratory Richland, WA 99352 Prepared for Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC FIN B2289 NUREG/CR-6181 PNL-9020
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Abstract As part of the Nondestructive Evaluation Reliability Program sponsored by the U.S. Nuclear Regulatory Commission, the Pacific Northwest Laboratory is dev-eloping a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components. This method uses probabilistic risk assessment (PRA) results and Failure Modes and Ef-fects Analysis (FEMA) techniques to identify and pri-oritize the most risk-important systems and components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot applications of this method. The specific systems addressed in this report are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results iii provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of im-proved inspection plans for nuclear power plants. To develop inspection plans, the acceptable level of risk
- from structural failure for important systems and com-ponents will be apportioned as a small fraction (i.e.,
5%) of the total PRA-estimated risk for core damage.
This process will deteI1nine target (acceptable) risk and target failure probability values for individual compo-nents. Inspection requirements will be set at levels to assure that acceptable failure probabilities are main-tained.
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I Contents Abstract............................................................................... iii Executive S1immary *..................................................................... vii Acknowledgments............................ *..........................................
xv Acronyms............................................................................. xvii Previous Reports in Series................................................................
xix 1.0 Introduction..................................................... *................... 1.1 2.0 Overall Methodology...................................... *............................ 2.1 2.1 Selection and Risk Prioritization.................................................... 2.1 2.2 Estimates of Component Rupture Probabilities......................................... 2.1 2.3 Target Risk iUld Rupture Probability................................................. 2.3 3.0 Analyses of Surry-1 Plant Systems........................................................ 31 31 Plant Familiarization............................................................ 31 3.1.1 Initial Plant Visit....................... *................................... 3.1
- 31.2 Information Obtained..................................................... 3.2 31.3 Subsequent Plant Visits.................................................... 3.2 31.4 Utility Interface........ *................................................. 3.2 3.2 Plant System Description......................................................... 3.2 3.21 React9r Pressure Vessel................................................... 3.2 3.2.2 Reactor Coolant System................................................... 3.3 3.2.3 Low-Pressure Injection System............................................... 3.5 3.2.4 Auxiliary Feedwater System................................................ 3.7 3.3 Analyses Assumptions........................................................... 3.9 3.4 Component Prioritization......................................................... 3.9 3.5 Results of Analyses............................................................ 310 3.6 Sensitivity and Uncertainty Analyses................................................ 311 3.61 Treatment of Uncertainties................................................ 3.11 3.6.2 Resul_ts of Uncertainty /Sensitivity Analyses.................................... 312 4.0 Discussions of the Results............................................................. 41 41 Ranking of Component Risk....................................................... 41 4.1.1 High-Risk Importance Components........................................... 41 41.2 Medium-Risk Importance Components........................................ 41 41.3 Low-Risk Importance Components
~.......................................... 4.2 4.2 Development of Target Risk and Rupture Probability Values............................... 4.3 5.0 Summary and Conclusions............................................................. 51 6.0 References......................................................................... 6.1 Appendix A: Sample of Component Importance Calculations...................................... Al V
NUREG /CR-6181
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Figures S.1.
Risk Contributions of Surry-1 Components for the Four Systems Addressed by this Study.............. xi S.2.
Risk Contributions of Surry-1 Components Based on Conditional Core Damage Given a Component Failure.......................................................................... xii S.3.
Cumulative Risk Contributions of Surry-1 Components...................................... xiii 2.1.
Information Provided to Expert Panel.......... *......................................... 2.3 2.2.
Estimates of Failure Probabilities for Surry-1 Reactor Pressure Vessel Components from Expert Judgement Elicitation............. *.......................................................,.. 2.4 3.1.
Surry-1 Reactor Pressure Vessel Simplified Schematic....................................... 3.3 3.2.
Surry-1 Reactor Coolant System Simplified Schematic.............................. _......... 3.4 3.3.
Surry-1 Low-Pressure Injection/Recirculation System Simplified Schematic....................... 3.6 3.4.
Surry-1 Accumulator System Simplified Schematic.......................... ;............... 3.7 3.5.
Surry-1 Auxiliary Feedwater System Simplified Schematic.................................... 3.8 3.6.
Risk Contributions ofSurry Components............................................... 3.1 3.7.
Cumulative Risk Contributions for Surry-1 Components.................................... 3.20 3.8.
Risk Contribution of Surry Components Based on Conditional Core Damage Given the Rupture...... 3.21 Tables S.1.
Risk Importance Parameter for Surry-1 Components........................................ ix 2.1.
System Levellmportance Ranking for Surry-1....................... *...................... 2.2 3.1.
Component Rankings Based on Core Damage Frequency for Four Selected Systems at Surry-1....... 3.13 3.2.
Component Rankings Based on Conditional Core Damage Frequency Given a Component Rupture(a) for Selected Systems at Surry-1......................................................... 3.15 3.3.
Risk Importance Parameters for Components at Selected Systems at Surry-1..................... 3.17 4.1.
Component Importance Compared with ASME BPVC Section XI Classifications and ISi Requirements for Selected Systems at Surry-1.............................................. *.......... 4.4 NUREG/CR-6181 vi
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Executive Summary As part of the Evaluation and Improvement of Nonde-structive Evaluation (NDE) Reliability for Inservice Inspection of Light Water Reactors Program sponsored by the U.S. Nuclear Regulatory Commission (NRC),
the Pacific Northwest Laboratory (PNL) has developed and applied a method using risk-based techniques to establish inservice inspection (ISi) plans for nuclear power plant components. As described in this report, the method uses probabilities of component failures
( estimated by using an expert judgment elicitation pro-cess) and plant-specific probabilistic risk assessment (PRA) results in conjunction with the Failure Modes and Effects Analysis (FMEA) technique to establish ISi priorities for systems and components. Included in this report is an approach for determining target risk and target rupture probability values for nuclear power plant components.
The Surry Nuclear Power Station Unit 1 (Surry-1) was selected for demonstrating the risk-based methodology.
The specific systems addressed in this report are the reactor pressure vessel, reactor coolant, low-pressure injection including accumulators, and the auxiliary feed-ater systems. The FMEAs were initially formulated sing plant system drawings and other plant-specific information. The Standard Review Plan information developed by the NRC was used in determining the effects of system failures. To ensure that the plant models were as realistic as possible, visits at the Surry-1 plant were conducted for plant system walkdowns and discussions were held with plant operational and techni-cal staff. Participation of Virginia Electric Power Com-pany staff was an essential part of the pilot study.
Because of similarities in objectives, the PNL program is coordinated with the American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection Guidelines. This task force has made general recommendations on the application of risk-based methods to inservice inspection and will make specific proposals to ASME for improved codes and standards. Results of PNL studies are being made available to the ASME group to demonstrate and vali-date the usefulness of the risk-based concepts.
The results of the risk-based component prioritization are summarized in Figure S.1. For the purpose of comparison, the components were also ranked on the basis of a conditional probability of core damage given a component failure, as shown by Figure S.2.
Vil Table S.1 shows the risk importance parameters for Surry-1 components. Included in the table are the estimated rupture probabilities for the components of the systems analyzed.
On the basis of core damage frequency, contributions of component failure to core damage frequency range from about 1.6E-14 to 1.58E-06 per plant year. The cumulative risk contribution for all of the components considered was estimated to be about 2.lE-06 per plant year. Figure S.3 shows the results of cumulative risk contribution for Surry-1 components within the systems analyzed. This estimate is about 5% of the total Surry PRA risk. The total estimated risk is dominated by failures of the reactor pressure vessel components (86%
of the total PRA risk). This risk is followed by the low-pressure injection system components (10%), and then other various components within the auxiliary feedwater and reactor coolant systems (4%). The results provide a guide to establish improved inspection priorities for nuclear power plant components.
To address uncertainties in the numerical results of the study, sensitivity analyses were performed. Based on uncertainties in estimated probabilities of component rupture probabilities, the sensitivity analyses results indicated no significant changes in component risk rankings (as shown in Figure S.1). Sensitivity analyses were also performed to determine overall core damage frequency due to component failures by indirect effects (pipe whip, jet impingement effects, etc.). The results indicate that contributions to the overall core damage frequency from the indirect effects were negligible.
Included in the report is a comparison of the risk-based inspection priorities suggested by this study with the current Surry-1 plant ISI practices. ASME classifica-tions and ISI requirements are generally in quantitative agreement with the risk-based rankings based on core damage frequency. The components making the great-est contribution to the core damage frequency have the most stringent ASME inspection requirements (i.e.,
both volumetric and surface examinations).
The analysis for the Surry-1 plant will be completed by developing the risk importance of components in the remaining systems (e.g., high-pressure injection, service water, and balance of plant). Similar plant-specific analyses will be performed for other pressurized-water reactors and for boiling-water reactors. Generic trends NUREG /CR-6181
Executive Summary in component importance will be established from these plant-specific evaluations. Once the high-priority com-ponents have been identified, recommended inspection programs (method, frequency, and extent) will be devel-oped. Probabilistic structural mechanics will be applied to establish inspection strategies that will ensure the component failure probabilities are maintained at ac-ceptable levels.
NUREG /CR-6181 Vlll A'
To develop inspection plans, the acceptable level of risk from structural failures will be apportioned as a small fraction of the total PRA-estimated risk for core dam-age. This process will determine target (acceptable) risk and target failure probability values for individual components. Inspection requirements will be set at levels to assure that acceptable failure probabilities are maintained.
If Executive Summary Table S.1. Risk Importance Parameter for Surry-1 Components Conditional Core Damage Frequency Rupture Core Dam-System-Component Rank Given Rup-Frequency age Fre-tore quency RPV - Beltline Region Welds 1
1.0 1.58E-06 1.58E-06 RPV - Beltline Plate 2
1.0
.1.00E-07
- 1.00E-07 RPV - Lower /Bottom Shell 3
1.0 7.32E-08 7.32E-08 AFW - CST, Supply Line 4
l.7E-02 4.03E-06 6.86E-08 RPV - Circumferential Flange to Nozzle Course 5
1.0 6.16E-08 6.16E-08 Upper Shell, Outside Beltline Welds LPI-A - Pipe Segment Between Accumulator Dis-6 1.8E-02 2.59E-06 4.67E-08 charge Header and RCS Isolation Valves LPI - Pipe Segment Between Containment Isolation 7
3.2E-02 1.30E-06 4.16E-08 Valve (inside) and Cold Leg Injection LPI - Pipe Segment Between Containment Isolation 8
3.20E-02 1.19E-06 3.80E-08 Valve (inside) and Hot Leg Injection LPI - LPI Sources (RWST, Sump), Supply Line 9
3.64E-02 1.00E-06 3.64E-08 LPI - Pipe Segment Between Pump Discharge and 10 3.2E-02 8.63E-07 2.76E-08 Containment Isolation Valve LPI - Pipe Segment Between Containment Isolation 11 1.6E-02 9.13E-07 1.46E-08 Valves RPV-CRDMs 12 5.0E-04 1.00E-05 5.00E-09 RPV - Instrument Lines 13 5.0E-04 1.00E-05 5.00E-09 AFW - Pipe Segment Between Containment Isolation 14 8.49E-05 3.92E-05 3.33E-09 and SG Isolation Valves AFW - Main Steam to AFW Pump Turbine Drive 15 1.64E-04 1.28E-05 2.lOE-09 RCS - Pipe Segment Between Loop Stop Valve and 16 1.13E-02 1.42E-07 1.60E-09 RPV (Cold Leg)
LPI - LPI Pump Suction Line 17 1.36E-03 1.lOE-06 1.50E-09 RCS - Pressurizer Spray Line 18 1.0E-04 1.00E-05 1.00E-09 RCS.- Pipe Segment Between RPV and Loop Stop 19 2.86E-03 2.00E-07 5.72E-10 Valve (Hot Leg) ix NUREG/CR-6181
Table S.1 (cont'd)
Executive Summary Conditional Core Damage Frequency Rupture Core Dam-System-Compon~nt Rank Given Rup-Frequency age Fre-ture quency AFW - AFW TD Pump Discharge Line 20 5.2E-05 1.02E-05 5.26E-10 AFW - Pipe Segment from Unit 2 AFW Pumps 21 1.4E-04 2.98E-06 4.18E-10 AFW - AFW Isolation Valve to SG 22 2.46E-06 6.SlE-05 1.60E-10 RPV - Nozzle to Vessel Welds 23 3.0E-03 2.00E-08 6.00E-11 RPV - Vessel Studs 24 5.0E-04
. 1.00E-07 5.00E-11 AFW - AFW MDP Suction Line 25 1.2E-05 3.55E-06 4.27E-11 AFW -AFW.MDP Discharge Line 26 1.65E-05 2.39E-06 3.95E-11 RPV - Upper, Closure Head, Flange 27 1.79E-03 2.00E-08 3.58E-11 RPV - Nozzle Forging Inlet/Outlet 28 1.25E-03 2.00E-08 2.50E-11 AFW - AFW TDP Suction Line 29 2.47E-06 6.12E-06 1.51E-11 AFW - Pipe Segment from Emergency Makeup 30 3.9E-06 1.46E-06 5.71E-12 System, Fire Main RCS - Pressurizer Relief/Safety Line 31 3.53E-07 6.14E-06 2.26E-12 RCS - Pressurizer Surge Line 32 1.5E-06 6.lE-07 9.15E-13 LPI-A - Accumulator Discharge Line 33 3.SE-08 2.0E-07 9.09E-14 RCS - Pipe Segment Between SG and RCP 34 3.05E-07 2.0E-07 6.lOE-14 RCS - Pipe Segment Between Loop Stop Valve and 35 3.0SE-07 1.41E-07 4.30E-14 SG (Hot Leg)
RCS - Pipe Segment Between RCP and Loop Stop 36 3.05E-07 7.75E-08 2.36E-14 Valve ( Cold Leg)
LPI-A - Accumulator, Suction Line 37 3.5E-08 4.57E-07 1.60E-14 NUREG/CR-6181 X
10-s 10-s 10-1 10-11
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10-12 10-13 10*14 0
1 RPV-Beltllne Region Welds 11 2
RPV*Bellllne P&ate 3
RPV-Lower/Bottom Shell 12 4
AFW-CST, Supply Line 13 5
RPV*Clr. Flange to Nozzle Course, 14 Upper Shell, Outside BeltUne Welds LPl*A-PJpe Segment BalwQ8n Acc.
15 Discharge Header & RCS Isolation Valves 16 7
LPI-Plpe Segmen! Between Containment lsola1lon Valve (Inside) & Cold Lag lnJOC1lon 17 LPI-Plpe Segment Betw&8n Containment lsol.
18 Valve (Inside) & Hot Leg Injection 19 9
LPJ,Source (RWST,Sump), Supply Line 10 LPl*Pfpe Segment Between Pump Discharge &
20 Containment lsol. Velva 21 5
10 LPI-Plpe Segment Between Containment 22 Isolation Valves 23 APV*CADMS 24 APV*lnstrument Lines 25 AFW*Plpe Segment Between Contalnmen1 26 Isolation & SG Isolation Valves
~
AFW-Maln Steam to AFW Pump Turbine Drive 28 RCS-Pipe Segment Between Loop Stop Vatve 29
& APV (Cold Leg) 30 LPl*LPI Pump Suction Line RCS-Pressurizer Spray Line 31 RCS-Pipe Segment Between RPV & Loop Stop 32 Valve(Hotleg) 33 AFW*AFW TD Pump Discharge Una 34 AFW-Pipe Segment from Unit AFW Pumps 35 15 25 and 75 { ::C Median Quartiles 1 Value AFW-lsolatlon Valve to SG RPV-Nozzle to Vessel Welds RPV-Vessel Studs AFW*AFW MOP Suction Line AFW-AFW MOP Discharge Une APV*Upper, Closure Head Range RPV-Nozzle Forging Outlet AFW*AFW TOP Suction Line AFW-Pipe Segment from Emergency Makeup System, Are Main RCS-Pressurizer ReHef/Satetyu Una RCS-Pressurizer Surge line LPl*A-Aecunwlator Discharge Uno RCS-Pipe Segment Between SG and RCP RCS-Pipe Segment Between Loop Stop Valve & SG (Hot leg) 20 25 36 37 2J}J ACS-Pipe Segmem Between 1 RCP & Loop Stop Vatve (Cokt Leg)
LPl*A-Accumulator, Suction Una 30 35 Component Identification R9111050.3 Figure S.1. Risk Contributions of Surry-I Components for the Four Systems Addressed by this Study
~:
101 10° 10-1 10*2 10-3 10-4 10*5 10-a 10-1 10-s 0
Risk-Based Rankings 2
3 9
4 8 10 6
5 7
11 18 19 26 29 12 16 31 14 28 30 27 13 20 34 33 15 17 32 36 21 22 35 23 24 25 37 Rani<
1 2
3 4
- 6 7
8 9
10 11 12 13 14 15 16 17 18 Syatem Component RPY*BeHllne Roglon*Weld*
RPV-BoHllno PlalO RPV-1..ower/Bottom Shall RPV..Clr. Flange to Nozzle Course, Upper Shell, Outside Beltllne Wekla
~Saun:H(RWST, Sumpl, Supply Uno LPI-Plpe Segment BetwNn Contalnmant IIOI. Valw (lnalde) I Cold Lag lnjectlon
~Plpo Segmont Be1Wftn Pump DIK!wgo & C...lolnmonl loo, Volvo LPI-A.Plpe Segment BatwNn Acc. Discharge Hnder & RCS laalatlon V.alvH AFW.CST, Suppl~ Uno
~Pipe Segment Bolweon ConlO!nmonl lsol. Volvo (ln11dol 6 Cold Log lnjoclJan LPf.Plpe Segment Between Contalnment lsol. VatveL RCS-Plpo Sogmonl Botweon Loop Slop Volvo & RPV (Cold l.ogl RPV-Nozzle to Vtnel1 Wekls RCS-Plpo Sogmonl Botween RPV & Loop Slop Volvo (Kol l.ogl RPY*Upper Cloaure Head Fl.Inge
~L.Pt Pump SocUon Uno RPV-No-Forging lnleUOullet RPV.CRDIIS 5
10 15 18 RPV-lnatrument Unn 20
- RPV*Vnael Studa 21 RCS.Preasurlzer!R*U*f Safety Un*
22 ACS-Pra1aurlzer Surge Line 23 RCS-Plpo Soglnonl Bolwoon SQ ond RCP 24 RCS-Plpo Segmonl a....... Loop Slop Volw & SQ (lfal l.ogl 25 RCS-Plpo Segmont a....... RCP & Loop Slop Volw (Cold Loal 28 AFW*Plpe Segment BetwNn* Containment taolallon Yaiva and SQ laollUon V*lvn n
AFW-lsolallon Valve ID SQ 28 AFW*AFW TD Pump DIKhorgo u..
21 AFW*llaln Stum to AFW Pump Turbine Drlve 38 AFW*Plpo Sogmonl fn,m Untt AFW Pumpo 31 RCS-Pnuurlz<< Spray Uno 32 AFW*AFWTDP Sue11'"1 Une 33 AFW*AFW IIDP Dlaclwgo U..
34
- AFW*AFWIIDP Sucllan Uno 35 LP~A-Accumulolor DIKlwgo Uno 38 AFW*Plpo Segmonl fn,m Ernorgoncy lluoup Sp,..., Flre lloln 37 LPI-A-Accumulator, Suction UM 20 25 30 35 Component Identification 40 R9111050.4 Figure S.2. Risk Contributions of Surry-I Components Based on Conditional Core Damage Given a Component Failure
I Executive Summary 2.0x 10*6 RPV
- Main Steam to AFW Pump 31 RCS
- Pressurizer Relief/Safety Line RPV
- Bellllne Plate Turbine Drive 32 RCS
- Pressurizer Surge Line RPV
- Lower/Bottom Shell 16 RCS
- Pipe Segment Between Loop 33 Lip-A. Accumulator Discharge Line AFW. CST, Supply Line Stop Valve and RPV (Ccld Leg) 34 RCS* Pipe Segment Between RPV *Cir.Flange to Nozzle Course, 17 LPI
- Pressurizer Spray Line 35 RCS
- Pipe Segment Between Loop 6
- A
- Pipe Segment Between Acc.
19 RCS. Pipe Segment Between RPV and Stop Valve and SG (Hot Leg)
Discharge Header and RCS Loop Stop Valve (Hot Leg) 36 RCS
- A
- Accumulator, SucHon Line Containment lsol. Valve (Inside) and AFWPumps Cold Leg Injection 22 AFW
- Isolation Valve to SG 8
- Pipe Segment Between 23 RPV
- Vessel Studs Hot Leg lnJecUon 25 AFW
- AFW MDP Suction Line 9
- Pipe Segment Between Pump 27 RPV. Upper, Closure Head, Flange Discharge and Containment lso. Valve 28 RPV
- Nozzle Forging Inlet/Outlet 11 LPI
- Pipe Segment Between Contain.
29 AFW
- Pipe Segment from Emergency 12 RPV*CRDMS Makeup System, Fire Main 13 RPV
- Instrument Lines 14 AFW
- Pipe Segment Between Containment Isolation and SG lsolatlon Valves OIL...--.....L..---11.....---...L..------L---J......-----'----.L.---.......... ___.
0 5
10 15 20 25 30 35 40 Component Identification R9108092.4 Figure S.3. Cumulative Risk Contributions of Surry-1 Components Xlll NUREG/CR-6181
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I Acknowledgments This work was supported by the U.S. Nuclear Regulato-ry Commission {NRC) under a Related Service Agree-ment with the U.S. Department of Energy under con-tract DE-AC06-76RLO 1830. The authors wish to acknowledge the direction and support provided by Dr.
Joe Muscara, NRC Program Manager. Dr. L. R.
Abramson from the NRC staff provided guidance to the elicitation process. Acknowledgments are also xv addressed to the many Virginia Electric Power Compa-ny staff for their participation in this work, particularly Ms. C. G. Lovett, Mr. R. K. MacManus, Mr. A.
McNeil, Mr. D. Rogers, Mr. D. Sommers, and Mr. E.
W. Throckmorton. The authors wish to thank T. T. Taylor, B. W. Smith, and T. W. Bardell for their contributions and review of this work.
AFW ASME ASTM BPVC BWR CRDM FMEA FSAR IRRAS ISI LOCA LPI LPI/LPR MDP MOY NDE NRC P&ID PNL PRA PWR RCP RCS RPV RWST SG TDP VEPCO VIMS Acronyms Auxiliary Feedwater System American Society of Mechanical Engineers American Society of Testing and Materials Boiler and Pressure Vessel Code Boiling Water Re~ctor Control Rod Driven Mechanisms Failure Modes and Effects Analysis Final Safety Analysis Report Integrated Reliability and Risk Analysis System Inservice Inspection Loss of Coolant Accident Low Pressure Injection System Low Pressure Injection/Recirculation System Motor Driven Pump Motor-Operated Valves Nondestructive Evaluation U.S. Nuclear Regulatory Commission Piping and Instrumentation Diagram Pacific Northwest Laboratory Probabilistic Risk Assessment Pressurized Water Reactor Reactor Coolant Pump Reactor Coolant System Reactor Pressure Vessel Reactor Water Storage Tank Steam Generator Turbine Driven Pump Virginia Electric Power Company Video Information Management System xvii NUREG/CR-6181
Previous Reports in Series
. Doctor, S. R., A. A. Dia7., J. R. Friley, M. S. Good, M. S. Greenwood, P. G. Heasler, R. L. Hockey, R. J.
Kurtz, F. A. Simonen, J.C. Spanner, T. T. Taylor, and T. V. Vo. 1993. Nondestructive Examination (NDE)
Reliability for Inservice Inspection of Light Water Reac-tors. NUREG/CR-4469, PNL-5711, Vol. 15. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., A. A. Dia7., J. R. Friley, M. S. Good, M. S. Greenwood, P. G. Heasler, R. L. Hockey, R. J.
Kurtz, F. A. Simonen, J.C. Spanner, T. T. Taylor, and T. V. Vo. 1992. Nondestructive Examination (NDE)
Reliability for Inservice Inspection of Light Water Reac-tors. NUREG/CR-4469, PNL-5711, Vol. 14. Pacific Northwest Laboratory, Richland, Washington.
Green, E. R., S. R. Doctor, R. L. Hockey, and A. A.
Diaz. 1992. Development of Equipment Parameter
- Tolerances for the Ultrasonic Inspection of Steel Compo-nents: Application to Components up to 3 Inches Thick.
NUREG/CR-5817, Vol 1. Pacific Northwest Laborato-ry, Richland, Washington.
Green, E. R., S. R. Doctor, J,l. L. Hockey, and A. A.
Diaz. The Interaction Matrix Study: Models and Equip-ment Sensitivity Studies for the Ultrasonic Inspection of.
Thin Wall Steel. NUREG/CR-5817. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., M. S. Good, P. G. Heasler, R. L. Hock-ey, F. A. Simonen, J. C. Spanner, T. T. Taylor, and T. V. Vo. 1992. Nondestructive Examination (NDE)
Reliability for Inservice Inspection of Light Water Reac-tors. NUREG/CR-4469, PNL-5711, Vol. 13. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., M. S. Good, P. G. Heasler, R. L. Hock-ey, F. A. Simonen, J. C. Spanner, T. T, Taylor, and T. V. Vo. 1992. Nondestructive Examination (NDE)
Reliability for Inservice Inspection of Light Water Reac-tors. NUREG/CR-4469, PNL-5711, Vol. 12. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., M. S. Good, E. R. Green, P. G. Heasler, F. A. Simonen, J. C. Spanner, T. T. Taylor, and T. V.
Vo. 1991. Nondestructive Examination (NDE) Reli-ability for Inservice Inspection of Light Water Reactors.
NUREG/CR-4469, PNL-5711, Vol. 11. Pacific North-west Laboratory, Richland, Washington.
xix Heasler, P. G., T. T. Taylor, J. C. Spanner, S. R. Doc-tor, and J. D. Deffenbaugh. 1990. Ultrasonic Inspection Reliability for Intergranular Stress Co"osion Cracks: A Round Robin Study of the Effects of Personnet Proce-dures, Equipment and Crack Characteristics.
NUREG/CR-4908. Pacific Northwest Laboratory, Richland, Washington.
Spanner, J. C., S. R. Doctor, T. T. Taylor/PNL and J.
Muscara/NRC. 1990. Qualification Process for Ultra-sonic Testing in Nuclear Inservice Inspection Applica-
- tions. NUREG/CR-4882, PNL-6179. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, T. T. Taylor, and T. V. Vo. 1990. Nondestructive Ex-amination (NDE) Reliability for Inservice Inspection of Light Water Reactors. NUREG/CR-4469, PNL-5711, Vol. 10. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, and-T. T. Taylor. 1989. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. NUREG/CR-4469, PNL-5711, Vol. 9. Pacif-ic Northwest Laboratory, Richland, Washington.
Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, and T. T; Taylor. 1989. Nondestructive Examination (NDE) Reliability_ for Inservice Inspection of Light Water Reactors'. NUREG/CR-4469, PNL-5711, Vol. 8. Pacif-ic Northwest Laboratory, Richland, Washington.
Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, and T. T. Taylor. 1988. Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors. NUREG/CR-4469, PNL-5711, Vol. 7. Pacif-ic Northwest Laboratory, Richland, Washington.
Doctor, S. R., J. D. De_ffenbaugh, M. S. Good, E. R.
Green, P. G. Heasler, G. A. Mart, F. A. Simonen, J.C.
Spanner, T. T. Taylor, and L. G. Van Fleet. 1987.
Nondestructive Examination (NDE) Reliability for Inser-vice Inspection of Light Water Reactors. NUREG/CR-4469, PNL-5711, Vol. 6. Pacific Northwest Laboratory, Richland, Washington.
Previous Reports Doctor, S. R., D. J. Bates, J. D. Deffenbaugh, M. S.
Good, P. G. Heasler, G. A. Mart, F. A. Simonen, J.C.
Spanner, T. T. Taylor, and L. G. Van Fleet. 1987.
Nondestructive Examination (NDE) Reliability for Inser-vice Inspection of Light Water Reactors. NUREG/CR-4469, PNL-5711, Vol. 5. Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., D. J. Bates, J. D. Deffenbaugh, M. S.
Good, P. G. Heasler, G. A. Mart, F. A. Simonen, J. C.
Spanner, A. S. Tabatabai, T. T. Taylor, and L. G. Van Fleet. 1987. Nondestructive Examination (NDE) Reli-ability for Inservice Inspection of Light Water Reactors.
NUREG/CR-4469, PNL-5711, Vol. 4. Pacific North-west Laboratory, Richland, Washington.
Collins, H. D. and R. P. Gribble.. 1986. Siamese Imag-ing Technique for Quasi-Veltical Type (QVT) Defects in Nuclear Reactor l'iping. NUREG/CR-4472, PNL-5717.
Pacific Northwest Laboratory, Richland, Washington.
Doctor, S. R., D. J. Bates, R. L. Bickford, L. A.
Charlot; J. D. Deffenbaugh, M. S. Good, P. G. Heasler, G. A. Mart, F. A. Simonen, J. 'C. Spanner, A. S.
Tabatabai, T. T. Taylor, and L. G. Van Fleet. 1986.
Nondestructive Examination (NDE) Reliability for Inser-vice Inspection of Light Water Reactors. NUREG/CR-.
4469, PNL-5711, Vol. 3. Pacific Northwest Laboratory, Richland, Washington; Doctor, S. R., D. J. Bates, L. A. Charlot, M. S. Good,.
H. R. Hartzog, P. G. Heasler, G. A. Mart, F. A.
Simonen, J.C. Spanner, A. S. Tabatabai, and T. T.
Taylor. 1986. Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Re-actors. NUREG/CR-4469, PNL-5711, Vol. 2. Pacific Northwesi Laboratory, Richland, Washington.
Doctor, S. R., D. J. Bates, L.A. Charlot, H. D. Collins, M. S. Good, H. R. Hartzog, P. G. Heasler, G. A. Mart, F. A. Simonen, J.C. Spanner, and T. T. Taylor. 1986.
Integration of Nondestructive Examination (NDE) Reli-ability and Fracture Mechanics, Semi-Annual Report,
- April 1984 -September 1984. NUREG/CR-4469, PNL-5711, Vol. 1. Pacific Northwest Laboratory, Richland, Washington.
NUREG/CR-6181 xx:
Good, M. S. and L. G. Van Fleet. 1986. Status of Activities for Inspecting Weld Overlaid Pipe Joints..
NUREG/CR-4484, PNL-5729. Pacific Northwest Labo-ratory, Richland, Washington.
Heasler, P. G., D. J. Bates, T. T. Taylor, and S. R.
Doctor. 1986. Perfonnance Demonstration Tests for Detection of Intergranular Stress Co"osion Cracking.
NUREG/CR-4464, PNL-5705, Pacific Northwest Labo-ratory, Richland, Washington.
Simonen, F. A. 1984. The Impact of Nondestructive Examination Unreliability* on Pressure Vessel Fracture Predictions. NUREG/CR-3743, PNL-5062. Pacific Northwest Laboratory, Richland, Washington.
- Simonen, F. A. and H. H. Woo. 1984. Analyses of the Impact of Inservice Inspection Using Piping Reliability Model. NUREG/CR-3869, PNL-5149. Pacific North-west Laboratory, Richland, Washington.
Taylor, T. T. 1984. An Evaluation of Manual Ultra-sonic Inspection of Cast Stainless Steel Piping.
NUREG/CR-3753, PNL-5070. Pacific Northwest Labo-ratory, Richland, Washington.
Bush, S. H. 1983. Reliability of Nondestructive Exami-nation, Volumes I, II, and III. NUREG/CR-3110-1, -2, and ~3; PNL-4584. Pacific Northwest Laboratory, Rich-
.land, Washington.
Simonen, F. A. and C. W. Goodrich. 1983. Parametric Calculations of Fatigue Crack Growth in Piping.
NUREG/CR-3059, PNL-4537. Pacific Northwest Labo-ratory, Richland, Washington.
Simonen, F. A., M. E. Mayfield, T. P. Forte, and D.
Jones. 1983. Crack Growth Evaluation for Small Cracks in Reactor-Coolant Piping. NUREG/CR-3176, PNL-4642. Pacific Northwest Laboratory, Richland, Washington.
Taylor, T. T., S. L. Crawford, S. R. Doctor, and G. J.
Posakony. 1983. Detection of Small-Sized Near-Surface Under-Clad Cracks for Reactor Pressure Vessels.
NUREG/CR-2878, PNL-4373. Pacific Northwest Labo-ratory, Richland, Washington.
Busse, L. J., F. L. Becker, R. E. Bowey, S. R. Doctor, R. P. Gribble, and G. J. Posakony. 1982. Characteriza-tion Methods for Ultrasonic Test Systems.
NUREG/CR-2264, PNL-4215. Pacific Northwest Labo-ratory, Richland, Washington.
Morris, C. J. and F. L. Becker. 1982. State-of-Practice Review of Ultrasonic In-service Inspection of Class I System Piping in Commercial Nuclear Power Plants.
NUREG/CR-2468, PNL-4026. Pacific Northwest Labo-ratory, Richland, Washington.
xxi Previous Reports Becker, F. L., S. R. Doctor, P. G. Heasler, C. J. Morris, S. G. Pitman, G. P. Selby, and F. A. Simonen. 1981; Integration of NDE Reliability and Fracture Mechanics, Phase I Report. NUREG/CR-1696-1, PNL-3469. Pacif-ic Northwest Laboratory, Richland, Washington.
Taylor, T. T. and G. P. Selby. 1981. Evaluation of ASME Section XI Reference Level Sensitivity for Initia-tion of Ultrasonic Inspection Examination.
NUREG/CR-1957, PNL-3692. Pacific Northwest Labo-ratory, Richland, Washington.
'I 1.0 Introduction Pacific Northwest Laboratory (PNL) is conducting a multi-year program for the U.S. Nuclear Regulatory Commission (NRC) entitled "Evaluation and Improve-ment in Nondestructive Evaluation Reliability for Inser-vice Inspection (ISi) of Light Water Reactors". The goals of this program are to determine the reliability of current ISi of pressure boundary systems and compo-nents, and to develop recommendations that. can ensure high inspection reliability. The long-term objective is to develop technical bases for improvements to the inspec-
- tion requirements of nuclear power plant components.
Because of similarities in objectives, the PNL program is coordinated with the American Society of Mechanical Engineers (ASME) Research Task Force on Risk-Based Inspection Guidelines. The initial task force document (ASME 1991) has made general recommen-dations on the application of risk-based methods to ISi, and forms the basis of future proposals to ASME for improved codes and standards. Results of PNL studies are being made available to the ASME group to dem-onstrate and validate the usefulness of the risk-based methodology. Future documents specifically addressing uclear. power plant components will be issued by the SME Task Force.
To provide technical bases for improved ISi plans, PNL has developed and applied a method (Vo et al. 1989) that uses results of probabilistic risk assessments (PRAs) to estimate the consequences of component failures. The probab~ties of these component failures have been estimated by using an expert judgment elici-tation process (Vo et al. 1991). Using these estimates of consequences and probabilities, risk calculations have established ISi priorities for systems and components at nuclear power plants. Once high-priority components have been identified, recommended inspection pro-grams (method, frequency, and extent) will be devel-oped in future work. Probabilistic structural mechanics will be applied to establish inspection strategies that will ensure that component failure rates are maintained at acceptable levels. After candidate inspection strategies yielding component failure probabilities less than identi-fied target values have been determined, decision analy~
sis techniques can be used to identify optimum inspec-tion strategies.
1.1 This report describes evaluations for the Surry Nuclear Power Station Unit 1 (Surry-1) which was selected for demonstrating the risk-based methodology. Participa-tion of Virginia Electric Power Company (VEPCO) staff was an essential part of the pilot study. Plant-specific information was obtained through system draw-ings, visits to the plant site, and discussions with plant operational staff. The specific systems selected for study were the reactor pressure vessel, reactor coolant, low-pressure injection including accumulators, and the auxiliary feedwater systems. The remaining pressure boundary systems at Surry-1 will be addressed in a future report. This report presents the results for the most risk~important components within the four select-ed systems at Surry-1 and compares the results for ISi priorities with the current ISi practices. Differences are being assessed to determine the extent of potential improvements to ISi plans provided by the new meth-odology.
Section 2.0 of this report discusses the overall method-ology for risk-based ranking of systems and com-ponents. Part of this discussion addresses the methods used to estimate component rupture probabilities. An approach is proposed for setting target values for these rupture probabilities at suitable levels. The objective of this approach is to ensure that the contribution of pres-sure boundary failures to core damage risk remains a small fraction of total plant risk.
Section 3.0 provides details of the Surry-1 pilot study.
Descriptions are provided for the four systems addressed, and the assumptions made in the analyses.
are also included. Results of the component rankings as well the sensitivity and uncertainty analyses are pre-sented. Section 4.0 provides a detailed discussion and interpretation of the results of Section 3.0. Finally, a summary and conclusions of the study are presented in
- Section 5.0.
2.0 Overall Methodology The overall methodology has three major steps: 1) selection and risk-prioritization of systems and compo-nents for inspection, 2) selection of a total target risk value associated with all pressure boundary and struc-tural failures, and 3) determination of target rupture probability for individual components or structures.
The following subsections summarize the overall meth-odology.
2.1 Selection and Risk Prioritization Both the Inspection Importance Measure (I~ devel-oped by PNL (Vo et al. 1989) and the Failure Modes and Effects Analysis (FMEA) technique were used to identify and prioritize the most risk-important systems and components for inspection. Previous work at PNL has addressed priorities at a system level as a prelude to component level prioritization. In summary, for a given system, Iw is defined as the product of the Birn-baum Importance (I8 ) times the failure probability for that system.
(2.1)
I8 = the change in risk that is associated with a system failure Pf = system failure probability due to struc-tural integrity failures.
Core-damage frequency (Level-I PRA) was used in this study as the bottom-line risk measure to prioritize the plant systems. When risk is measured by core damage frequency, the I8 of a system is equivalent to the condi-tional probability of core damage given a system failure.
The parameter I8 is a measure of the consequence of structural failure, where Iw also addresses the proba-bilities of structural failures. Specifically, Iw is an ap-proximation of the core damage risk due to system failures.caused by structural failures. Components with very high I8 values are given the highest rank in the risk-based prioritization for inspection planning. Com-ponents having high Iw values are then added into this list of highly ranked components. The results of system prioritization for Surry-1 from an earlier PNL study (Vo et al. 1990) are shown in Table 2.1.
For those systems selected for further analyses, a de-tailed component-level prioritization was performed.
-he FMEA teclmique was selected fo, this analysis.
2:1 The FMEA results were used.to calculate the impor-tance index ( or relative importance) for each compo-nent within the selected systems. The importance index was based on the expected consequence of failure of the component, as measured by the probability of core damage resulting from component failure. In mathe-matical terms, the probability of core damage, Pcm*
resulting from a given component failure (i.e., rupture),
is defined as where Pcm
= probability of core damage resulting from the component failure
= failure (rupture) probability of the component of interest conditional probability of core dam-age given the failure of system i conditional probability of system i failing given the component failure
= probability that the operator fails to recover given a system failure.
Equation (2.2) is totaled for all system failures that either result directly from the given component failure or result indirectly from component damage to other compone~ts or the systems in the zone of interest ( e.g.,
pipe whip or jet impingement effects, damaging vital electrical buses, etc.). The Standard Review Plan devel-oped by the NRC (1981) and information obtained from plant system walkdowns are used to assess the indirect effects.
As shown by Equation (2.2), estimates of component failure probabilities are required in order to perform component prioritization. These estimates are summa-rized in the following subsection.
2.2 Estimates of Component Rupture Probabilities For each system selected (Table 2.1), the per-compo-nent failure probability was estimated. Because histori-cal failure data on low-probability events ( e.g., pipe rupture) are lacking, an expert judgment elicitation was used to estimate component failure probabilities. This section summarizes the procedures and the results of NUREG/CR-6181
2.0 Overall Methodology Table 2.1. System Level Importance Ranking for Surry-l(a)
System 1W (IB)
Ranking High-Pressure Injection 1.3E-05 (1.4E-02) 1 (3)
Low-Pressure Injection 6.lE-06 (1.6E-02) 2 (2)
Reactor Pressure Vessel 5.0E-06 (1.0) 3 (1)
Auxiliary Feedwater 3.9E-07 (8.2E-03) 4 (4)
Service Wate/b) 1.0E-07 (2.2E-03) 5 (5)
Steam Generator 5.lE-08 (5.lE-06) 6 (8)
Reactor Coolant 2.9E-08 (6.lE-04) 7 (6)
Power Conversion 1.9E-09 (5.lE-06) 8 (7)
(a)
Obtained from Vo et al. (1990). Values in parentheses represent the system Birnbaum Importance Measure results or their associated rankings.
(b)
Including the component cooling water system.
PNL's expert judgment elicitation. More detailed dis-cussions are given in Vo et al. (1991).
The expert judgment elicitation used a systematic pro-cedure, which closely followed the approaches reported in the NRC Severe Accident Risks Document (NRC 1989; Wheeler et al. 1989; Meyer et al. 1989). The specific objective of the PNL elicitation was to develop numerical estimates for probabilities of catastrophic or disruptive failures in the selected components at Surry-
- 1. Figure 2.1 shows information that was used to ob-tain the* desired estimates from the experts.
Prior to the expert elicitation workshop, PNL sent reference materials to the experts, including data sourc-es, reports, probabilistic models, and recent PRA re-sults. Panel members were asked to study these mate-rials and to make initial estimates of failure probabil-ities.
At the meeting, a formal presentation was provided for each system addressed. Presentations covered technical descriptions, historical component failure mechanisms, elicitation statements, suggested approaches, question-naire forms, and any materials that supported the issue descriptions. The presentations were followed by dis-cussions. The experts provided their knowledge regard-ing plant design and operation, failure history, material degradation mechanisms, and methods for recomposi-.
tion and aggregation of the data.
NUREG/CR-6181 2.2 Each expert then completed questionnaire forms that addressed location-specific rupture probabilities for the systems of interest. These responses included best estimates of probabilities and uncertainties, and the rationale for these estimates. Following the meetin-,
the information provided by the expert panel was re composed and aggregated. PNL prepared a prelim' report of the elicitation, which was then submitted to each panel member for review. This report included the initial recomposition, additional plant-specific data, and other relevant information. The experts were requested to review and revise their estimates of rup-ture probabilities. The revised information was again recomposed and aggregated to provide single composite judgments for each issue.
Figure 2.2 shows a sample of estimated failure probabil-ities obtained from the expert judgment approach.
Similar types of plots were produced for components in other selected systems at Surry-1. For readability, the probabilities are presented with a log10 scale, with the probabilities expressed as failures per component per year. The ranges of best estimates from the experts were summarized in a series of plots (boxes and whis-kers) as shown in Figure 2.2. An individual plot dis-plays five features of the distribution of estimated prob-abilities. The "whiskers" display the extreme upper and lower bound values of the distribution, while the box itself locates the 25% and 75% quartiles of the distribu-tion. Finally, the circle within the box is the median of the distribution.
2.0 Overall Methodology Data from PRA Results and Historical Fracture Mechanics Other Relevant Information Failure Data Analyses (system, component prioritization, system descriptions, etc.)
~,
~,
~,
Expert Judgment Additional Information Elicitation and (additional plant-specific Discussion information, etc.)
~,
Estimated Rupture Probabilities Figure 2.1. Information Provided to Expert Panel 2.3 Target Risk and Rupture Proba-bility The purpose of inspections is to keep risk levels within acceptable values ( e.g., by detecting and repairing de-graded components before they lead to rupture). It is a difficult task to select acceptable ( or target) values for the risk associated with pressure boundary component and structural failures. However, once the target risk values are determined, the corresponding target rup-ture probabilities for individual components can be quantified using the methods described in the preceding subsections. This subsection proposes an approach for defining the target risk and the target rupture probabili-ties.
A philosophical approach for selecting target values of risk and rupture probability for individual components has been recommended by ASME Research Task Force on Risk-Based Inspection (ASME 1991). This approach assumes that the inspection should ensure 2.3 that the risk of core damage resulting from pressure boundary component structural failures is maintained to be less than a small fraction of the total core damage risk estimated by the PRA. The risk due to pressure boundary structural failures is referred to as the "target risk" and 5% of the total PRA-estimated risk resulting from internal events has been recommended as an appropriate numerical value.
It is further recommended that this overall target risk be apportioned among the risk-important components by considering the risk associated with rupture of each component. Using the conditional probability of core damage given component failure, the target failure probability can be calculated for each component from its apportioned share of the overall target risk value.
From this, inspection strategies can be determined to maintain component rupture probabilities below target values, and optimum strategies can then be selected.
An example of the target risk and component rupture probability calculations is provided in the next section to clarify the discussion.
2.0 Overall Methodology Weld 1 Weld2 Weld3 Weld4 Weld5 Weld6 Weld7 Weld8 Weld9 Weld 10 Weld 11 Weld 12 Weld 13 Weld 14 Weld 15 Nozzle Forgings - Inlet Nozzle Forgings - Outlet Beltline Plate Vessel Shell - Outside Beltline Upper Head Lower Head Vessel Flange Enclosure Head Flange Vessel Studs CRDMs Instrument Line Penetrations
~
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~
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~
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Legend:
Weld 1 -
Circumferential weld, upper shell to intermediate shell Weld 2 -
Circumferential weld, flange to nozzle Weld 3 -
Circumferential weld, lower shell (beltline region)
Welds 4 -
Circumferential welds, thru 5 bottom head Welds 6 -
Intermediate and lower shell thru 9 longitudinal welds (beltline regions)
Welds *1 O -
Nozzle to vessel welds thru 15 CRDMs -
Control rod drive mechanisms Upper/Lower Bound
...._ _ _.l 25%175% Quartile
<j>
Median
~
<p
~ ~
~
I I
I I
I I
I
-9
-8
-7
--6
-5
-4
-3 log 10 (FailuresNear)
S9012060.2 Figure 2.2. Estimates of Failure Probabilities for Surry-1 Reactor Pressure Vessel Components from Expert Judgement Elicitation NUREG/CR-6181 2.4
3.0 Analyses of Surry-I Plant Systems This section presents analyses of the four selected Surry-1 systems. Identification and prioritization ~f components for the Surry-1 plant systems are provided following a brief discussion of plant familiarization, system descriptions, and analysis assumptions. The section concludes with sensitivity analyses.
3.1 Plant Familiarization Participation of VEPCO was an essential part of the pilot study. Before initiating the pilot study, a visit to VEPCO headquarters was conducted. The purpose of this first visit was to get acquainted with VEPCO per-sonnel and to request needed data.
Prior to the initial plant visit, the project team analysts reviewed the fault trees reported in the Surry-1 PRA, the system descriptions, and the sections of the final safety analysis report (FSAR) applicable to the systems of interest. The preliminary FMEA models were con-structed and preliminary success criteria and dependen-cy matrices were developed to identify specific areas here information was needed to develop an accurate odel. Based on these initial activities, a letter of quest was prepared and sent to the plant to identify the plant-specific information and data that was required. The following subsections provide a descrip-tion of the plant visit and the information obtained during the visit.
3.1.1 Initial Plant Visit A one-week plant visit was arranged to meet with plant personnel. During this visit, project team analysts performed the system walkdowns and obtained relevant plant information. The visiting PNL team included plant system specialists and PRA specialists. Because the plant was in operation during the initial visit, system walkdowns for some locations were not possible ( e.g.,
inside the containment building and other high-radia-tion areas). Therefore, the Video Information Manage-ment System (VIMS) developed by VEPCO was also used. VIMS is a computerized system, that displays photographs of plant systems and components that have been stored in digital form on a laser disc. Following simple instructions, the plant photographs could be retrieved and viewed at any location within the plant.
For each of the systems selected for the study, a system alkdown was conducted where possible. The informa-3.1 tion obtained from the walkdowns was later used to assess the indirect effects on the systems. The walk-downs for each system included the plant engineer and one or two project team analysts. For each component (e.g., pipe segment), all the necessary information relat-ed to that component was obtained. This information was entered into the preliminary FMEA niodels. For example, for a given pipe segment within a selected system, the component identification, including the pipe size, W<!,S identified. Ntimbers of welds, elbows, sup-ports, connections, penetrations, etc., within the pipe segment in question were identified and recorded.
Given a component failure, the potential targets that might be impacted by the failed components ( e.g., vital electrical buses, system components nearby, etc.) were also recorded. Additionally, a video camera was used to record the conversations with the responsible engi-neer and views of significant locations of concern to sys-tem design and operation as identified during the initial visit.
In addition to the plant system walkdowns, discussions with plant operational and technical staff were also conducted. The areas of discussion included plant and system modeling questions, collections of system design and operational information, discussions of transient sequence progressions, and the operators' responses to these events. During the plant visit the team had dis-cussions with the Surry-1 supervisor of system safety, the operator training coordinator, and the supervisor of the ISi. Project analysts talked with reactor operators, the shift technical advisor, and members of the mainte-nance and engineering staff.
Discussions centered on gaining a clear understanding of the following items:
the normal and emergency configurations and operations of the various systems of interest system dependencies operational problem areas identified by plant personnel that may impact the analysis automatic and manual actions taken in re-sponse to various emergency conditions availability of plant specific operational data.
3.0 Analyses The emergency procedures which addressed actions identified by the project analysts as important actions were explained to operations personnel.
3.1.2 Information Obtained A complete set of the current Surry piping and instru-mentation drawings (P&ID), isometric drawings, com-posite drawings, and stress analysis reports were provid-ed by the Surry-1 staff. Also, the Surry-1 staff provided copies of the Surry Emergency Procedures, Abnormal Procedures, Emergency Contingency Action Proce-dures, Functional Restoration Procedures, and several sections from the current revisions of the Surry-1 FSAR. The plant information was incorporated into PNL's preliminary FMEA models. For instance, the isometric and composite drawings were used to obtain additional information regarding component orientation and number of subcomponents. The Emergency Proce-dures were used to assess the recovery actions by the operators given a rupture of component.
3.1.3 Subsequent Plant Visits During the course of the study, two additional plant visits were conducted. The first visit was to obtain ad-ditional plant-specific failure mechanisms for compo-nents within the system analyzed. This information was provided to an expert workshop on estimating compo-nent rupture probabilities. The other plant visit was conducted during the plant shutdown for refueling.
This visit was to obtain additional information and to verify the information that was obtained from an initial visit ( e.g., areas inside the containment building). PNL is currently performing the pilot study for the Surry-1 balance of the plant system and additional plant visits are anticipated.
3.1.4 Utility Interface An ongoing interface was maintained with the utility throughout the duration of the analysis. The project team leader was in frequent contact with Surry-1 plant personnel to ask questions and verify information.
Surry-1 personnel also reviewed the results of the study when they became available.
NUREG /CR-6181 3.2 3.2 Plant System Description Surry-1 is part of a two-unit plant located on the James River near Williamsburg, Virginia. Surry-1 is a West-inghouse-designed, three-loop, pressurized-water reactor (PWR) rated at 788 MWe capacity with a sub-atmo-spheric containment. The balance of the plant and containment building were designed and constructed by Stone and Webster Engineering Corporation. Surry-1 is operated by VEPCO. Commercial operation started in 1972.
The Surry-1 systems selected for study were the primary pressure boundary system, the front-line safety systems, and certain important support systems identified in Table 2.1. These Were the reactor pressure vessel (RPV), reactor coolant (RCS), low-pressure injection (LPI), and the auxiliary feedwater (AFW) systems. The following paragraphs summarize the descriptions for these systems. Detailed descriptions can be found in the Surry-1 FSAR.
3.2.1 Reactor Pressure Vessel The RPV is a principal component of the RCS.
Surry-1 RPV is shown in Figure 3.1. It consists of a cylindrical shell with a hemispherical bottom head, and a flanged and gasketed removable upper head. The vessel contains the core, core support structures, control rQds, thermal shield, and other parts directly associated with the core. Outlet and inlet nozzles are located between the upper head and the core.
The Surry-1 vessel was designed and manufactured by the Babcock and Wilcox Company to the requirements of Section Ill of the ASME Boiler and Pressure Vessel Code (BPVC). Design features and materials selection are typical for PWR reactor vessels at U.S. nuclear plants. The vessel is designed of low-alloy steel with forgings of Type A508, Class 2 and plate materials of Type A533, Grade B, Class 1. All surfaces in contact with coolant are clad with, or made from, 300-series stainless steel or Inconel. In general, all attachments and pressure-containing parts have full-penetration welds. Partial welds are used to attach the relatively small diameter control rod drives and the instrumenta-tion tubes to the vessel heads.
3.0 Analyses Flange Ligaments 1 thru 58 (Ref, 1-1100A) 2 _...,.....t---.-------,r-------l~
Diameter: 157.0" I.D.
Circumference: 492.98" I ~
6 (45°}
3--..-
I 8~
(135°}
9 (315°}
4--..-
7 (225°}
Material: Flange-A508 Class 2 Carbon Steel Upper Shell: 9.125"T - 508 Class 2 Carbon Steel Intermediate Shell: 9.0"T -
A533 Carbon Steel Lower Shell: 9.0"T - A533 Carbon Steel Bottom Shell: 5.375"T - A30 Class 2 Carbon Steel Bottom Head: 5.375"T - A533 Carbon Steel Welds 6,7,8 & 9: 100" Length 89201018.1 Figure 3.1. Surry-1 Reactor Pressure Vessel Simplified Schematic 3.2.2 Reactor Coolant System The function of the RCS is to remove heat and transfer it to the secondary system. It also provides a barrier against the release of reactor coolant or radioactive materials to the containment environment. ' The RCS for Surry-1 is diagramed in Figure 3.2. It consists of 3.3 three identical heat transfer loops ( connecting parallel to the RPV), each of which includes a steam generator, reactor coolant pump, and connecting piping and instru-mentation for flow and temperature measurements.
The pipes through which the heated water flows from the RPV to the steam generator are called the "hot legs" and the pipes through which the cooled water NUREG/CR-6181
3.0 Analyses LOOP C
£t.:TOR ODLANT UMP DETAIL OF~~~~~---
CONNECTION \\~
- 19.
Figure 3.2. Surry-1 Reactor Coolant System Simplified Schematic flows from the steam generator and back into the RPV are called the "cold legs." The working fluid is boiled on the secondary sides of the steam generator and transported through a conventional turbine-condenser system.
The RCS also includes a pressurizer that maintains the reactor coolant at a constant pressure. The pressurizer system consists of power-operated relief valves with associated block valves, ASME code safety valves, pres-surizer sprays, and electrical heaters. There is continu-ous control of the water and steam inventory within the pressurizer vessel. The pressurizer is connected to a NUREG /CR-6181 3.4 coolant loop and is maintained at the saturation tem-perature that corresponds to the system pressure.
To regulate the reactor coolant chemistry within design limits and control the pressure level, a constant letdown flow from one loop upstream of the reactor coolant pump is maintained. This flow is, in turn, controlled by the pressurizer level. Constant coolant makeup is add-ed by charging pumps in the chemical and volume control systems. The inservice integrity of the RCS is addressed through periodic inspections performed in accordance with the requirements of ASME,Section XI.
3.2.3 Low-Pressure Injection System
. The LPI consists of several independent subsystems characterized by equipment and flow path redundancy inside the missile protection boundaries. The two phas-es of low-pressure system operation including active low-pressure injectiqn and recirculation mode and the passive accumulator injection are summarized below.
The Surry-1 low-pressure injection/recirculation system (LPI/LPR) provides emergency coolant injection and recirculation following a loss-of-coolant accident (LOCA) when the RCS depressurizes below the low-pressure setpoint (about 300 psig). In addition to the direct recirculation of coolant during the recirculation phase once the RCS is depressurized, the LPR dis-charge provides the suction source for the high-pressure recirculation system following drainage of the refueling water storage tank (RWST).
- The LPI/LPR at Surry-1 is diagrammed in Figure.3.3.
The system consists of two 100% capacity pump trains.
In the injection mode, the pump trains share a common
. operated valves (MOVs), check valves, and locked-open manual valves. Each pump discharges through a check valve and normally open MOV in series to a common injection header. The injection header contains a locked-open MOV and branches to separate lines, one to* each cold leg. Each of the lines to the cold legs contains two check valves in series to provide isolation from the high-pressure RCS.
In the recirculation mode, the pump trairis draw suction from the containment sump through a parallel arrange-ment of suction lines to a common header. Flow froni the suction header is drawn through a normally closed MOV and check valve in series. Discharge of the pump is directed to either the cold legs through the same lines used for injection or to a parallel set of headers that feed the charging puinps, depending on the RCS pressure.
In the hot-leg injection mode, system operation is iden-tical to normal recirculation with the exception that the normally open cold-leg injection valves must be manual-ly closed remotely, and one or more normally closed hot-leg recirculation valves must be manually opened.
3.5 3.0 Analyses The accumulators, which are passive components, serve as another injection mode for the LPI system.. They provide an initial influx of borated water to reflood the reactor core following a large or medium LOCA. The accumulator system, diagrammed in Figure 3.4, consists of three tanks filled with borated water and pressurized with nitrogen. Each of the accumulators is connected to one of the RCS cold legs by a line containing*a normally open MOV and check valve in series. The check valves serve as isolation valves during normal operation and open to empty the contents of the accu-mulators when the RCS pressure falls below 650 psig.
The accumulators depend on the nitrogen system to maintain the pressure head. The nitrogen is supplied by dedicated local nitrogen bottles, and the accumula-tors are fully instrumented to indicate abnormal pres-sure conditions. The accumulators are initially filled with borated water storage from the RWST, and the valves are dosed. Instrumentation verifies that the level remains above a minimum value.
The associated components, piping, structures, and power supplies of the LPI system (including the accu-mulators) are designed to conform with Class 1 seismic criteria. All motors, instruments, transmitters, and their associated cables located inside the containment are designed to function during and under the postulated temperature, pressure, and humidity conditions.
All LPI piping in contact with borated water is austenit-ic stainless steel. The piping is designed to meet the minimum requirements set forth in B31.l Code for Pressure Piping, B36.10 and B16.19, ASTM Standards, Supplementary Standards, and Additional Quality Con-trol Measures. The piping is supported to accommo-date expansion due to temperature changes and hydrau-lic forces during an accident. All components of the LPI/LPR and accumulators are tested periodically to demonstrate system readiness. All pressure piping butt welds containing radioactive fluid, at greater than 600°F and 600 psig, were radiographed. The remaining piping butt welds were randomly radiographed. Pressure-containing components are inspected for leaks from pump seals, valve packing, flanged joints, and safety valves during system testing. Frequency of testing and maintenance of the system components are specified in the ASME,Section XI.
Sump To HPI XV48 PS33 (1-SI-P-18)
MDPSIIB (1-SI-P-1 A)
PS32 MDPSIIA To Charging Pump Inlet Header NC-FAI CV50 NOFAI 1864D PS35 PS34 NOFAI D 1864A 1863A PS39 1890B CV228 Power Removed NOFAI 6"-Sl-152-1502 1890C PS36 1890A CV229 PS46 From HPI 6"-Sl-50-1502 From HPI CV241 PS44 CV242 PS45 6"-Sl-49-1502 CV243 From HPI To Charging Pumps AL - Out of Position Alarm in Control Room CV79 CV82 CV85 R9312053.3 Figure 3.3. Surry-1 Low-Pressure Injection/Recirculation System Simplified Schematic Hot Leg Loop 3 Hot Leg Loop 2 Cold Leg Loop 1 Cold Leg Loop 2 Cold Leg Loop 3 Hot Leg Loop 1
1-S1-TK-1A From RWST FC 1865A 1-SI-TK-1B FC 1865B 1-S1-.TK-1C
.FC 1865C CV107 3.0 Analyses Loop 1 CV 1
~
0 9----1 Cold Leg Loop2
,.1-----....;r,1------1 Cold Leg CV128 CV130 CV145 CV147 R9312053.1 Figure 3.4. Surry-1 Accumulator System Simplified Schematic 3.2.4 Auxiliary Feedwater System The AFW system provides feedwater to the steam generators for heat removal from the primary system after a reactor trip. The AFW system may also be used following a reactor shutdown, in conjunction with the condenser dump valves or atmospheric relief valves, to cool the RCS to about 300°F and 300 psig, at which time the residual heat removal system is brought into operation. The AFW system also provides emergency water following a secondary-side line rupture. Removal of heat in this manner prevents the reactor coolant pressure from increasing and causing release of reactor coolant through the pressurizer relief and/or safety valves.
The AFW system is diagramed in Figure 3.5. The AFW is a multiple-train system; it consists of electric motor-driven pumps and steam turbine-driven pumps.
Each pump draws suction through an independent line from the condensate storage tank. Each.(',FW pump 3.7 discharges to parallel headers; each of these headers can provide AFW flow to any or all of the steam gener-ators. Flow from each header to any one steam gener-ator is through a normally open MOV and locked-open valve in series, paralleled with a line from the other header. These lines* feed one line containing a check valve that joins the main feedwater line to a steam generator.
The motor-driven pumps automatically start on receipt of a safety actuation system signal, loss of main feed-water, low steam generator level in any steam genera-tor, or Joss of off-site power. The turbine-driven pumps automatically start on indication of a low steam genera-tor level in any steam generator or undervoltage of any of the main RCS pumps.
Most of the AFW equipment is located in the auxiliary building. This building is designed to withstand the effects of earthquakes, tornadoes, floods, and other natural phenomenon. Provisions are incorporated in NUREG/CR-6181
w bo u,
0
'<t LO
'<t
(.)
X X
LO 300,000 GAL CST Header B Header A To Unit 2 AFW System PS94 MOVFW260A
.......... ~..... MOVFW260B XN'ZTO XV271 PS84 CV133 CV131 PS83 CV138
..,.,.... From Fire Main CV309
-..............,""""-""'"'16-""' From Emergency Makeup System
_______..,.._ Main Steam MOVFW160B XN87 XV120 PS96 ADVMS102B PS95 CV178 CV182 ADVMS102A Turbine Drive for PumpTDPFW2 From Unit 2 AFW Pumps
,J--..... ~~-
CV273 MOVFW160A R9312053.2 Figure 3.5. Surry-1 Auxiliary Feedwater System Simplified Schematic
the AFW design to allow periodic operation to dem-onstrate performance and structural leak-tight integrity.
Leak detection is provided by visual examination and sensors in the floor drain system. The capability to isolate components or piping is provided, if required, so that the AFW system's safety function will not be com-promised. Provisions are made to allow for ISi of components at. the appropriate times specified in the ASME,Section XI.
3.3 Analyses Assumptions General assumptions used for the analyses are the following:
Core damage frequency was used as the bottom-line risk measure to prioritize plant system components.
For the four selected systems, the discrete components* (piping segments, welds, etc.)
are identified for purposes of the risk-based evaluation. For the RPV, the major compo-nents of interest were the vessel shell, heads, flanges, closure studs, penetrations, nozzles and safe ends, and attachment welds. For other systems, the coinponents of interest were pipe segments. These included the straight lengths of pipe, pipe elbows, cou-plings, fittings, flanged joints, and welds.
Additionally, tanks and heat exchangers, including the pressurizer, are also included as components in the analyses.
The system Birnbaum Importance results were used to provide the conditional proba-bilities of core damage given the system failures.
Identical components in identical trains within the same system were assumed to have the same failure consequences.
In these analyses, failures in piping of less than 1-in. in diameter generally are not considered, primarily because of the enorm-ous amount of instrumentation piping of this.
size. Active functions of components such as pumps and valves, which make up part of 3.9 3.0 Analyses the system* pressure boundary, are not con-sidered. However, failures of these compo-nents as pressure boundaries are addressed.
. Steam generator tube failures have been considered in other studies and are not in-cluded in this study.
The Standard Review Plan 3.6.2, developed by the
-NRC (1981), was used in determining the indirect effects (e.g., pipe whip, jet forces, etc.) of compo-nent failures, as such failures relate to other com-ponents in the zone of interest ( e.g., vital electri-cal buses). Additionally, when a larger diameter pipe impacts a smaller diameter pipe of the same pipe schedule, a smaller diameter pipe is assumed to fail.
Potential flooding due to pipe ruptures that could damage safety-related systems and equipment are not included in these analy-ses. Floodings will be addressed at the later date.
3.4 Component Prioritization The quantitative FMEA technique,. as described in Section 2.0, was used to prioritize components on the basis of core damage risk. In summary, the following sources of information were used to prioritize compo-nents for inspection: 1) the component failure proba-bilities estimated from expert judgment elicitation (Vo et al. 1990), 2) the results froni Surry-1 system prioriti-zation (Vo et al. 1989), and 3) system fault trees report-ed in the Surry-1 PRA (Bertucio and Julius 1990). The
- Integrated Reliability and Risk Analysis System (IRRAS) computer program developed by Idaho Na-tional Engineering Laboratory (Russel et al. 1987) was used to reanalyze the developed fault trees ( e.g., calcu-late the conditional probability of system failure given a component failure).
The FMEAs were initially formulated using plant sys-tem drawings and other relevant plant-specific infor-mation. As stated in the assumptions, Standard Review Plan information developed by the NRC was used in determining the potential effects of system component failures on other components in the zone of interest.
To ensure that plant models were as realistic as possi-ble and reflected plant operational practices, visits to NUREG/CR-6181
3.0 Analyses the Surry-1 plant were conducted for plant system walk-downs, and discussions were held with plant operational and technical staff. For locations where the walkdowns were not possible, (e.g., high-radiation areas) the VIMS developed by VEPCO was used to identify the poten-tially impacted systems and equipment (given a failure of a component in the zone of interest).
The FMEA worksheets were devised so that the neces-sary information could be systematically tabulated. In the following paragraphs, the example of the RPV (using Equation 2.2) is discussed. Copies of FMEA worksheets are provided in Appendix A of this report.
The first step of the analysis was to identify the compo-nent locations and/or the number of subcomponents within a specified pipe segment or region. For exam-ple, the beltline region of the Surry-1 RPV consists of five welds (four longitudinal welds and one circumfer-ential weld). The per-weld failure probability, Pr, was estimated as 3.2E-07 (see Figure 2.2). The failure probability. of the beltline region (five welds) was esti-mated as 5* (3.2E-07 /weld) = 1.6E-06.
As discussed in a previous section, the consequences of component failures were to be placed into two catego-ries, those that resulted in direct effects on the system in question and those that resulted in effects on other systems or components in the zone of interest ( e.g.,
component failures due to pipe whip or jet impinge-ment effects). In either case, the total contribution to core damage, given a failure of the component under consideration, were assessed ( e.g., the product of the conditional probability of core dam*age given system failures and the probability of system failures given component failures, Pcm Is
- P 6 IP).
Information from prior system level prioritizations, system walkdowns, discussions with VEPCO staff, the Standard Review Plan, and the fault trees reported in the Surry-1 PRA were used to quantify the failure effects. The system fault tree was reanalyzed to esti-mate the probability of system failure given a compo-nent failure. The IRRAS computer program was used to calculate probabilities of system failure given a com-ponent failure. Generally, the Birnbaum Importance Measure for the system was used to provide the condi-tional probability of core damage given a system failure.
For the RPV, the primary effect of a weld failure was the loss of the vessel (P s IP = 1.0). In this case, the NUREG/CR-6181 3.10 probability of core damage, given the vessel failure, Pcm I s, was assumed to be 1.0.
Depending on failure location and/or accident scenario, the recovery action, Ri, was assigned an estimated probability based on discussions with the plant technical staff and on information obtained from the PRA. For the reactor pressure, no recovery action was possible.
It is important to note that in this study the probability of recovery by the operator staff was incorporated in the system prioritization scheme. To prevent double counting, the recovery actions were assessed qualitative-ly. For each postulated failure of a component within the selected systems, the core damage probability was calculated. A computer program was developed for the calculations.
The product of the component failure probability and the corresponding core damage probability given a failure of the component, was calculated. This value describes the expected risk-based implication of the component under consideration. In the vessel example, failures of welds at the vessel beltline region were as-sumed to result in loss of the vessel, and the core dam-age probability per plant year was estimated to be (1.6E-06)
- 1 = i.6E-06.
On the FMEA worksheets the relative importance of each component was calculated as illustrated above (e.g., 1.6E-06
- 1). An importance index was used to rank each component in a given system by normalizing its core damage probabilities to that of a component with the highest core damage probability. The highest value of the index identifies the component that is the most important for the system being analyzed. A final combined ranking of components for all four systems together was developed based on the numerical values for core damage frequency.
3.5 Results of Analyses Within the four systems analyzed, there are approxi-mately 250 major components (or pipe segments). By assuming that identical components in identical trains within the same system have the same failure probabili-.
ties and consequences, these components are reduced to approximately 125 components. For ranking pur-pose, components within the same train can be further grouped, based on major discontinuities (e.g., between
- pumps and major valves). This resulted in 37 major component groups within the systems analyzed.
Table 3.1 shows the results of the risk-based ranking of major components within the four selected systems at Surry-1, based on the contributions of component fail-ures to core damage frequency. Included in the table are the upper-and lower-bound values estimated for each component to indicate the effects of uncertainties in the estimates of component rupture probabilities.
The rankings ( as shown in the table) are based on the median values estimated from the Surry-1 PRA and PNL evaluations of other factors such as rupture proba-bilities, as discussed in the preceding section. Fig-ure 3.5 presents this information graphically for the components in the four systems.
As shown in Table 3.1, the contributions of individual component failures to core damage frequency (based on the median values) range widely from about 1.6E-14 to 1.58E-06 per plant year. The cumulative risk contribu-tion from all components as shown in Figure 3.6 is about 2.lE-06 per plant year. Figure 3.7 shows this
- cumulative risk contribution. It is interesting to note that the risk contribution is domina:ted by approximately the first 18 highest-ranked components. The system level rankings obtained by summing component contri-butions are the following: 1) RPV, 2) LPI, 3) AFW, and 4) RCS. These system level rankings agree with those obtained in an earlier PNL study (Vo et al. 1989).
For the purpose of comparison, the components are also ranked on the basis of the.calculated values of
- conditional probability of core damage given a compo-nent failure. Table 3.2 shows this ranking based on the median values estimated. Figure 3.8 presents this same information graphically. The conditional contributions of component failures to core damage range from 1.0 to about 1.0E-07. As expected, the highest contributions are from RPV components, since rupture of the RPV beltline region leads directly to core damage. Condi-tional contributions to core damage from the LPI, the AFW, and the RCS components are lower due to the ability of redundant safety systems to mitigate accidents and, hence, prevent core damage. Table 3.3 shows the risk importance parameters for 37 major components identified in Table 3.1. The component rupture fre-quencies ( as shown in Table 3.3) were the average of component group.
3.11 3.0 Analyses 3.6 Sensitivity and Uncertainty Analy-ses There are various sources of uncertainty in the numeri-cal results of this study. This section describes specific sources of uncertainty.and provides the results of uncer-tainty/ sensitivity analyses.
3.6.1 Treatment of Uncertainties Two basic types of uncertainties addressed in this study were parameter value uncertainty and modelqig uncer-tainty. Parameter value uncertainties were evaluated for component rupture probabilities, the conditional probability of core damage* given component failures, and human recovery action probabilities. Modeling uncertainty was evaluated for the treatment of the indi-rect effects of the component failures.
The uncertainties of the component rupt~re probabili-ties have been addressed in Vo et al. 1990. For exam-ple, the population quartile was chosen to describe uncertainty in the estimates of component rupture probabilities (see Figure 2.2). Limited uncertainty analyses regarding the core damage conditional proba-bilities have been addressed. The uncertainties in com-ponent unavailabilities, initiating event frequencies, and cut set element unavailabilities and their associated modeling were not addressed in this study. Consider-ation of functional dependencies and common-cause effects on systems were based on the results evaluated by the selected PRAs. The mean parameter values estimated by the PRAs were used to calculate the core damage conditional probabilities. The uncertainties of recovery action errors were addressed in the Surry-1 PRA (Bertucio and Julius 1990). In these evaluations the probabilities were assessed using values of each parameter such that the nth percentile ( or quartile) of the uncertainty distribution representing the range over which the true values lie. In this, Monte Carlo or other sampling techniques were used to assess the propaga-tion of parameter value uncertainty through the final results.
There are many variables involved in calculating the indirect effects given a component failure ( e.g;, location of pipe break, orientation of the equipment, direction of whipping pipe, number of hangers and/ or supports, impact location, angle of impacts, etc.). Guidance NUREG /CR-6181
3.0 Analyses provided in the Standard Review Plan 3.6.2 and infor-mation obtained through discussions with VEPCO staff during system walkdowns were used to assess the indi-rect effects. The assessments of the indirect effects using Standard Review Plan 3.6.2 are likely to be con-servative. The uncertainty was evaluated by excluding the potential indirect effects of component failures in the model (e.g., pipe whip or jet impingement effects) and recalculating the overall core damage frequency.
3.6.2 Results of Uncertainty /Sensitivity Analyses Sensitivity analyses were performed on issues that could potentially have significant impact on component rank-ing. The sensitivity analyses addressed the changes in component rankings by using upper-and lower-estimat-ed values of component rupture probabilities as report-ed in calculations in Vo et al. (1990). As shown in Table 3.1, although variation exists in the numerical results, most components have relatively the sanie rank-ing, as compared to the ranking based on the median values. The largest variations in component ranking were the LPI supply lines and sources, pipe segments extending from isolation valves to the steam generator, and pump suction and discharge lines of the AFW system. Pipe segments between the RPV and the RCS.
loop stop valves have moderate variations in ranking.
Sensitivity analyses were also performed by letting the
- component rupture probabilities approach 1.0. This causes Pcm values to be the same as Pcm IP s** The new rankings are shown in Table 3.2 (and FigurJ 3.6). As shown in the table, the RPV components remain on top of the important-component list. This is followed by components within the LPI system, AFW system, and the RCS. The components that had the largest increas-es in ranking were the LPI source and supply line, ranking second, which is an increase of eight in compo-nent ranking. Other components remained at relatively the same rankings. Although probability of failure of the LPI suction line/source is quite low, loss of this component ( e.g., loss of common suction lines of the LPI supply line) could disable the entire system, there-by contributing significantly to core damage.
Sensitivity analyses were performed to address contribu-tions to core damage from indirect effects of compo-nent failures. The results show that contributions of NUREG/CR-6181 the indirect effects to the overall core damage frequen-cy are negligible (less than 2%). Two pipe segments identified to have potential failure effects on the other systems nearby were 1) the pipe segment between the accumulator discharge line and RCS isolation valve and
- 2) the pipe segment between LPI pump discharge line and the containment isolation valve. Rupture of the accumulator discharge line could result in a failure of the entire residual heat removal discharge line due to its potential pipe whip and/or jet impingement effects.
Similarly, rupture of the LPI pump discharge line could result in disabling of the charging pump inlet header.
Although analyses regarding potential flooding within the plant due to pipe ruptures were not part of this study, the safeguard room at Surry-1 needs to be men-tioned. The safeguard room houses the AFW pumps and serves as a pass-through area for the main steam
- lines and the main feedwater lines. The room also contains three main steam isolation valves, three main steam valves, fifteen steam generator atmospheric dump valves, three steam generator relief valves, one small decay heat relief valve, and three main feedwater check
- valves. The concern is that a rupture in a pipe seg-ment, a valve body, or a steam-water line could flood
- the room with steam and/or water, thereby causing all AFW pumps and other safety-related equipment to fail.
This is an important issue and will be addressed at the
- later date. Component risk prioritization of the entire reactor system will be completed and the main steam and main feedwater lines will also be evaluated at a later date.
3.12
Table 3.1. Component Rankings Based on Core Damage Frequency for Four Selected Systems at Surry-l(a)
System - Component(b)
Core Damage Frequency Upper Median Lower RPV-Beltline Welds 4.00E-04 1.58E-06 5.40E-07 RPV-Beltline Plate 1.00E-07 1.00E-07 8.00E-08 RPV~
Lower /Bottom Shell 2.SOE-07 7.32E-08 2.00E-08 AFW-CST, Supply Line 1.60E-07 6.86E-08 1.60E-08 RPV-Circumferential Flange to Nozzle 1.30E-07 6.16E-08 2.00E-08 Course, Upper Shell, Outside Beltline Welds LPI-A -
Pipe Segment Between Accumulator i.60E-07 4.67E-08 3.20E-09 Discharge Header and RCS Isolation Valves LPI -
Pipe Segment Between Containment 1.60E-07 4.16E-08 3.2E-08 Isolation Valve (inside) and Cold Leg Injection LPI -
Pipe Segment Between Containment 8.70E-08 3.80E-08 1.40E-08 Isolation Valve (inside) and Cold Leg Injection LPI -
LPI Sources (RWST, Sump), Supply 1.86E-07 3.64E-08 3.20E-10 Line LPI-Pipe Segment Between Pump Dis-8.64E-08 2.76E-08 8.80E-09 charge and Containment Isolation Valve LPI -
Pipe Segment Between Containment 5;44E-08 1.46E-08 8.00E-09.
Isolation Valves RPV-CRDMs 2.74E-08 5.00E-09 2.23E-09 RPV-Instrument Lines 4.18E-08 5.00E-09 1.58:E-09 AFW-Pipe Segment Between Containment 8.00E-09 3.33E-09 1.52E-10 Isolation and SG Isolation Valves AFW-Main Steam to AFW Pump Turbine 8.36E-09 2.lOE-09 2.79E-10 Drive RCS -
Pipe Segment Between Loop Stop 2.70E-09 1.60E-09 5.34E-10 Valve and RPV (Cold Leg) 3.0 Analyses Rank(c) 1 2
3 4
5 6
7 8
9 10 11 12 12 14 15 6
3.13 NUREG/CR-6181
3.0 Analyses Table 3.1 (cont'd)
System - Component(b)
Core Damage Frequency Rank(c)
Upper Median Lower LPI -
LPI Pump Suction Line 1.59E-09 l.SOE-09 4.48E-11 17 RCS -
Pressurizer Spray Line 1.60E-09 1.00E-09 4.70E-19 18 RCS -
Pipe Segment Between RPV and Loop 2.00E-09 5.72E-10 3.20E-11 19 Stop Valve (Hot Leg)
AFW-AFW TD Pump Discharge Line 6.00E-10 5.26E-10 2,48E-11 20 AFW-Pipe Segment from Unit 2 AFW 1.39E-09 4.18E-10 1.39E-10 21 Pumps AFW-AFW Isolation Valve to SG 2.60E-10 1.60E-10 4.70E-12 22 RPV-Nozzle to Vessel Welds 3.90E-10 6.00E-11 1.14E-11 23 RPV-Vessel Studs.
5.00E-10 5.00E-11 1.58E-11 24 AFW-AFW MDP Suction Line 1.39E-09 4.27E-11 2.79E-11 25 AFW-AFW MDP Discharge Line 5.41E-11 3.95E~ll 4.92E-12 26 RPV-Upper, Closure Head, Flange 5.71E-11 3.58E-11 5.71E-12 27 RPV-Nozzle Forging Inlet/Outlet 2.SOE-11 2.SOE-11 5.00E-12 28 AFW-AFW TDP Suction. Line 4.37E-11 1.SlE-11 5.41E-12 29 AFW-Pipe Segment from Emergency Make-2.46E-11 5.71E-12 2.46E-12 30 up System, Fire Main RCS -
Pres.surizer Relief/Safety Line 7.00E-12 2.26E-12 2.35E-13 31 RCS-Pressurizer Surge Line 2.23E-12 9.lSE-13 3.0SE-13 32 LPI-A -
Accumulator Discharge Line 5.56E-13 9.09E-14 3.47E-14 33 RCS-Pipe Segment Between SG and RCP 3.0SE-13 6.lOE-14 1.83E-14 34 RCS -
Pipe Segment Between Loop Stop 2.44E-13 4.30E-14 1.83E~14 35 Valve and SG (Hot Leg)
RCS -
Pipe Segment Between RCP and Loop 1.52E-13 2.36E-14 6.lOE-15 36 Stop Valve (Cold Leg)
LPI-A -
Accumulator, Suction Line 3.47E-14 1.60E-14 3.47E-15 37 (a)
- Based on the estimated median values of component rupture probabilities.
(b)
RPV = Reactor Pressure Vessel; AFW = Auxiliary Feedwater; LPI = Low Pressure Injection; LPI-A = Low Pressure Injection-Accumulator; RCS = Reactor Coolant System.
(c)
Rankings were based on "Median" values.
NUREG /CR-6181 3.14
RPV-RPV-RPV-AFW-.
RPV-LPI-A -
- LPI -
PI -
LPI -
LPI -
LPI -
RPV-RPV-AFW-AFW-RCS -
LPI -
RCS-RCS -
AFW-AFW-Table 3.2. Component Rankings Based on Conditional Core Damage Frequency Given a Component Rupture<a) for Selected Systems at Surry-1 System-Component(b)
Core Damage Frequency Beltline Welds 1.0 Beltline Plate 1.0 Lower /Bottom Shell 1.0 CST, Supply Line l.70E-02 Circumferential Flange to Nozzle Course, Upper Shell, 1.0 Outside Beltline Welds Pipe Segment Between Accumulator Discharge Header and l.SOE-02 RCS Isolation Valves Pipe Segment Between Containment Isolation Valve (inside) l.60E-02 and Cold-Leg Injection Pipe Segment Eetween Containment Isolation Valve (inside) 3.00E-02 and Hot-Leg Injection LPI Sources (RWST, Sump), Supply Line 3.64E-02 Pipe Segment Between Pump Discharge and Containment 2.00E-02 Iso. Valve Pipe Segment Between Containment Isolation Valves l.20E-02 CRDMs 5.00E-04 Instrument Lines 5.00E-04 Pipe Segment Between Containment Isolation and SG Isola-4.00E-04 tion Valves Main Steam to AFW Pump Turbine Drive l.60E-04 Pipe Segment Between Loop Stop Valve and RPV (Cold l.lOE-02 Leg)
LPI Pump Suction Line l.40E-03 Pressi.µ-izer Spray Line l.OOE-04 Pipe Segment Between RPV and Loop Stop Valve (Hot Leg) 2.90E-03 AFW TD Pump Discharge Line 2.30E-04 Pipe Segment from Unit 2 AFW Pumps l.40E-04 3.15 3.0 Analyses Rank 1
1 1
9 1
8 10 6
5 7
11 19 19 27 31 12 17 18 14 30 32 NUREG /CR-6181
3.0 Analyses Table 3.2 (cont'd)
System-ComponentCb)
Core Damage Frequency Rank AFW-AFW Isolation Valve to SG 2.60E-04 29
/
RPV-Nozzle to Vessel Welds 3.00E-03 13 RPV-Vessel Studs 5.00E-04 19 AFW-AFW MDP Suction Line 1.20E-05 34 AFW-AFW MDP Discharge Line 1.70E-05 33 RPV-Upper, Closure Head, Flange 1.SOE-03 15 RPV-Nozzle Forging Inlet/Outlet 1.30E-03 16 AFW-AFW TDP Suction Line 4.00E-05 28 AFW-Pipe Segment from Emergency Makeup System, Fire Main 3.90E-06 36 RCS -
Pressurizer Relief/Safety Line 5.00E-04 19 RCS -
Pressurizer Surge Line 5.00E-04 19 LPI-A -
Accumulator Discharge Line 1.00E-05 33 RCS -
Pipe Segment Between SG and RCP 5.00E-04 19 RCS -
Pipe Segment Between Loop Stop Valve and SG (Hot Leg) 5.00E-04 19 RCS -
Pipe Segment Between RCP and Loop Stop Valve (Cold 5.00E-04 19 Leg)
LPI-A -
Accumulator, Suction Line 2.lOE-07 37 (a)
Based on the estimated median values of component rupture probabilities.
(b)
RPV = Reactor Pressure Vessel; AFW = Auxiliary Feedwater; LPI = Low Pressure Injection; LPI-A = Low Pressure Injection-Accumulator; RCS = Reactor Coolant System.
NUREG /CR-6181 3.16
Table 3.3. Risk Importance Parameters for Components at Selected Systems at Surry-1(a)
Conditional Core Damage System-Component(b)
Frequency Rupture Rank Given Rupture Frequency RPV - Beltline Region Welds 1
1.0 1.58E-06 RPV - Beltline Plate 2
1.0 1.00E-07 RPV - Lower /Bottom Shell 3
. 1.0 7.32E-08 AFW - CST, Supply Line 4
1.7E-02 4.03E-06 RPV - Circumferential Flange to Nozzle 5
1.0 6.16E-08 Course Upper Shell, Outside Beltline Welds LPI-A - Pipe Segment Between Accumu-6 1.8E-02
- 2.s9E-06 lator Discharge Header and RCS Isola-tion Valves PI - Pipe Segment Between Contain-7 3.2E-02 1.30E-06 ment Isolation Valve (inside) and Cold Leg Injection LPI - Pipe Segment Between Contain-8 3.20E-02 1.19E-05 ment Isolation Valve (inside) and Hot Leg Injection LPI - LPI Sources (RWST, Sump), Sup-9 3.64E-02 1.00E-06 ply Line LPI - Pipe Segment Between Pump Dis-10 3.2E-02 8.63E-07 charge and Containment Isolation Valve LPI - Pipe Segment Between Contain-11 1.6E-92 9.13E-07 ment Isolation Valves RPV-CRDMs 12 5.0E-04 1.00E-05 RPV - Instrument Lines 13 5.0E-04 1.00E-05 AFW - Pipe Segment Between Contain-14 8.49E-05 3.92E-05 ment Isolation and SG Isolation Valves AFW - Main Steam to AFW Pump Tur-15 1.64E-04 1.28E-05 bine Drive RCS~ Pipe Segment Between Loop Stop 16 1.13E-02 1.42E-07 Valve and RPV ( Cold Leg) 3.17 3.0 Analyses Core Damage Frequency l.58E-06 1.00E-07 7.32E-08 6.86E-08 6.16E-08 4.67E-08 4.16E-08 3.80E-08 3.64E-08 2.76E-08 l.46E-08 5.00E-09 5.00E-09 3.33E-09 2.lOE-09 1.60E-09 NUREG/CR-6181
3.0 Analyses Table 3.3 (cont'd)
Conditional Core Damage Core System-Component(b)
Frequency Rupture Damage Rank Given Rupture Frequency Frequency LPI - LPI Pump Suction Line 17 1.36E-03 l.10E006 1.50E-09 RCS - Pressurizer Spray Line 18 1.0E-04 1.00E-05 1.00E-09 RCS - Pipe Segment Between RPV and 19 2.86E-03
- 2.00E-07 5.72E-10 Loop Stop Valve (Hot Leg)
AFW - AFW TD Pump Discharge Line 20 5.2E-05 1.02E-05 5.26E-10 AFW - Pipe Segment from Unit 2 AFW 21 1.4E-04 2.98E-06 4.18E-10 Pumps AFW - AFW Isolation Valve to SG 22 2.46E-06 6.SlE-05 1.60E-10 RPV - Nozzle to Vessel Welds 23 3.0E-03 2.00E-08 6.00E-11 RPV - Vessel Studs 24 5.0E-04 1.00E-07 5.00E-11 AFW - AFW MDP Suction Line 25 1.2E-05 3.55E-06 4.27E-11 AFW - AFW MDP Discharge Line 26 1.65E-05 2.39E-06 3.95E-11 RPV - Upper, Closure.Head, Flange 27 1.79E-03 2.00E-08 3.58E-11 RPV - Nozzle Forging Inlet/Outlet 28 1.25E-03 2.ooE~os 2.50E-11 AFW - AFW TDP Suction Line 29 2.47E-06 6.12E-06 1.51E-11 AFW. - Pipe Segment from Emergency 30 3.9E-06 1.46E-06 5.71E-12 Makeup System, Fire Main RCS - Pressurizer Relief/Safety Line 31 3.53E-07 6.14E~06 2.26E-12 RCS - Pressurizer Surge Line 32 1.SE-06 6.lE-07 9.15E-13 LPI-A - Accumulator Discharge Line 33 3.5E-08 2.0E-07 9.09E-14 RCS - Pipe Segment Between SG and 34 3.05E-07 2.0E-07 6.lOE-14 RCP RCS - Pipe Segment Between Loop Stop 35 3.0SE-07 1.41E-07 4.30E-14 Valve and SG (Hot Leg)
RCS - Pipe Segment Between RCP and 36 3.05E-07 7.75E-08 2.36E-14 Loop Stop Valve (Cold Leg)
LPI-A - Accumulator, Suction Line 37 3.SE-08 4.57E-07 1.60E-14 (a)
Based on the estimated median values.
(b)
RPV = Reactor Pressure Vessel; AFW = Auxiliary Feedwater; LPI = Low Pressure Injection; LPI-A = Low Pressure Injection-Accumulator; RCS = Reactor Coolant System.
NUREG /CR-6181 3.18
u C
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. 'S G) w
.Q C,
- - ca
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0 0
10*5 10*6 10*1 10-s 10*9
.10-10 10-11 10-12 10-13 1 o-14 0
1 2
3 4
5 9
10 RPV*Bsltline Region Welds RPV*Beltllne Plate RPV*Lower/Bottom SheU AFW*CST, Supp/'y Line RPV-Clr. Flange to Nozzle Course, Upper Shen. Outside Beltllne Welds LPl*A*Plpe Segmont &tween Acc.
Discharge Header & RCS Isolation Valves LPI-Plpe Segment Between Containment Isolation Valvo (lnslde) & Cold Leg Injection LPl*Plpe Segment Between Containment !sol.
Valve (Inside) & Hot leg Injection LPl*Sourco (RWST,Sump), Supply Lino LPl*Plpe Segment B9tween Pump Discharge &
Containment lsol. Valve 5
11 12 13 14 LPl*Plpe Segment Between Containment Isolation Valves RPV,CRDMS RPV*lnstrument Unes AFW*Plpe Segment Between Containment Isolation & SO Isolation Valves 15 AFW-Maln Steam to AFW Pump Turbine Drive 16
. RCS*Pfpe Segment Between Loop Stop Valve 17 18 19 20 21 10
& RPV (Cokl Log)
LPI-LPI Pu"lJ SUctlon Une RCS-Pressurizer Spray Line RCS*Plpe Segment Betwoen RPV & Loop Stop Volvo(Hot Log)
AFW*AFW TD Pump Otscharge Une AFW-Pipe Segment from Unit AFW PL1f11JS 15 22 23 24 25 26 27 28 29 30 31 32 33 34 35 25 and 75 { A Median Quartiles 1 Value AFW-lsolatlon Va!ve lo SG RPV-Nazzfe to Vessel Wekts RPV-Vessel Studs AFW-AFW MOP SUctlon Uno AFW-AFW MOP Discharge Line RPV*Upper, Closure Head Flange RPV,Nozzlo Forging OUtlOI AFW-AFW TOP SUctlon Una AFW-Plpe Segment from Emergency Makeup System, Fire Main RCS-Pressurizer ReMef/Safat)'U Una RCS-Pre&surizer surge Una LPl*A* Aca,uwlator Discharge Une RCS-Pipe Segment Betwaen SG and RCP RCS-Pipe Segment Between Loop S!cp Valve & SO (Hot Leg) 20 25 36 37 2J}J RCS-Pipe Segmortt Bo1woon
~
RCP & Loop Slop Volvo (Cold Log)
LPl*A*Ac:cumutator, Suction Uno 30 35 Component Identification R9111050.3 Figure 3.6. Risk Contributions of Surry Components w
b f
3.0 Analyses 2.0x 10*6 RPV - Beltllne Region Welds 15 AFW - Main Steam to AFW Pump 31 RCS - Pressurizer Relief/Safety Line RPV - Beltllne Plate Turbine Drive 32 RCS - Pressurizer Surge Line RPV - Lower/Bottom Shell 16 RCS - Pipe Segment Between Loop 33 Lip-A - Accumulator Discharge Line AFW
- CST, Supply Line Stop Valve and RPV (Cold Leg) 34 RCS-Pipe Segment Between RPV - Cir. Flange to Nozzle Course, 17 LPI - LPI Pump Suction Line SGand RCP Upper Shell, Outside Beltllne Welds 18 RCS - Pressurizer Spray Line 35 RCS* Pipe Segment Between Loop 6
LPI - A - Pipe Segment Between Acc.
19 RCS - Pipe Segment Between RPV and Stop Valve and SG (Hot Leg)
Discharge Header and RCS Loop Stop Valve (Hal Leg) 36
. RCS - Pipe Segment Between RCP Isolation Valves 20 AFW - AFW TD Pump Discharge Line and Loop Stop Valve (Cold Leg) 7 LPI - Pipe Segment Between 21 AFW - Pipe Segment from Unit 2 37 LPI - A - Accumulator, Suction Line Containment lsol. Valve (Inside) and AFWPumps Cold Leg lnJectlon 22 AFW - Isolation Valve to SG 8
- Pipe Segment.Between 23 RPV
- 9 LPI
- Pipe Segment Between Pump 27 RPV - Upper, Closure Head, Flange Discharge and Containment !so. Valve 28 RPV - Nozzle Forging Inlet/Outlet 11 LP!
- Pipe Segment Between Contain.
29 AFW - AFW TDP Suction Line Isolation Valves 30 AFW - Pipe Segment from Emergency 12 RPV-CRDMS Makeup System, Fire Main 13 RPV
- Instrument Lines 14 AFW - Pipe Segment Between Containment Isolation and SG Isolation Valves Component Identification R9108092.4
- Figure 3.7. Cumulative Risk Contributions for Surry-1 Components NUREG/CR-6181 3.20
101 10° 10-1 10-2 e u C
0 Cl) 0
- s 10-3 ca CT C l!!
ou.
Ea, 10-4 "C C>
C ca oE o ca Q
10-5 10-s 10-1 10-a 0
Risk-Based Rankings 2
3 9
4 8 10 6 5 7
11 18 19 26 29 12 16 31 14 28 30 27 13 20 34 33 15 17 32 36 21 22 35 23 24 25 37 Rank 1
- 3 4
5 8
7 8
9 10 11 12 13 14 15 18 17 18 Syatem Component RPV*B1ttllm Region-Welda RPV-8.tlllne Plate RPV*Lower/Bottom Shell RPY-Clr. Flange to Nozzle Course, Upper Shell. OUtalde Bettllne Wolds LPI-Soun:a (RWST, Sump~ Supply Uno LPI-Plpe s.an,ent BetwNn Containment laol. v,rv. (ln11d1J
- Cold Ltg lnjectlon LPI-Plpe Segment Between Pump DlKharge & Containment lso. Valve LPI-A*Pfpe Segment Betwnn Acc. Dlac:harge Headtr & RCS lsolaUon ValvH AFW-CST, Supply Uno LPI-Plpe Segment Between Containment laol. Valve (lnald1) & Cold Leg ln)ectlon LPI-Plpe Sogment Between Containment lsoL V1lvo1.
RCS-Pl po Segment lletwHn Loop Slop Volvo & RPV (Cold Leg)
RPV.ffozzl1 to Venell Welda RCS-Pl po Segment lletwHn RPV & Loop Slap Volvo (Hot Leg)
RPV-Uppor era...., Hood Flange
~LPI Pump Section Una RPV-Nozzlo Forging lnlol/Oullot RPV-CRDMS 5
10 15
,a RPV.Jnatrument Unn 20 RPV*Vnul stucb 21 RC&Pr.. 1urlztrlR1tlef Safety Une 22 RCS-Preasurtzor Surge Un, 23 RCS-Pipe Segment Between SO and ACP 24 RCS-Pipe Sogmonl Bolwnn Loop Slop Volvo & SO (Hot Log) 25 RCS-Pipe Segment Botwnn RCP & Loop Slop Volvo (Cold Log) 26 AFW*Plpo Segment Betwwn Containment Isolation Valve and SQ lsollltlon Yalvn 27 AFW*ltolatlon Valvo to SO 28 AFW-AFW TD Pump Dlschfllrgo Une 29 AFW-lllln Stum to AFW Pump Turbine Drlvo 30 AFW*Plpe Segment from UnH MW Pumpt 31 RCS-Preasurtzer Spray Uno 32 AFW*AFWTDP Suction Uno 33 AFW*AFW UDP Dl,chorgo Uno 34 AFW*AFW UDP Suction Uno 35 LPI-A*Accumulalor Discharge Une 38 AFW*Plpo Segment from Emergency Makeup 5yalem, Flre Main
- n LPI-A*Accumulalar, Suction Une 20 25 30 35 Component Identification 40 R9111050.4 Figure 3.8. Risk Contribution of Surry Components Based on Conditional Core Damage Given the Rupture w
b
4.0 Discussions of the Results This section discusses the results presented in Sec-tion 3.0. Proposed values of target risk and rupture probability are also discussed. The discussions are based on the estimated median parameter values.
4.1 Ranking of Component Risk The rankings of Table 3.1 were developed on the basis of core damage frequency. In this discussion we will identify. the factors that govern these rankings, begin-ning with the highest ranked components and ending with the lowest ranked components.
For discussion purposes, "high-risk importance compo-nents" refer to components that have core damage frequency between l.OOE-08 and > l.OOE-06. Similarly, "medium-risk importance components" refer to those components having core damage frequencies of 1.00E-10 to l.OOE-08. "Low-risk importance components" refer to those components having core damage frequen-cies less than l.OOE-10.
4.1.1 High-Risk Importance Components he most risk-important components are located within the beltline region of the reactor vessel. The impor-tance of this region is due to high neutron fluence and associated embrittlement levels, in conjunction with the high consequence ( core damage) resulting from struc-tural failure. The critical welds were identified on the basis of concerns with pressurized thermal shock for the Surry-1 vessels (Heinecke et al. 1987). Given a cata-strophic failure within the beltline region, core damage is certain. The lower and bottom shell portions of the reactor vessel were also important but had somewhat lower rankings due to the lower estimated rupture probabilities of these components ( e.g., lower ftuence and embrittlement levels). The circumferential flange-to-nozzle course welds, upper shell welds, and welds outside the beltline region of the pressure vessel had still lower rankings but were identified to be important.
Although these components have lower neutron fluence and lower enibrittlement levels, failures of these compo-nents also have high core damage consequences.
High-risk importance (approaching RPV components) for the AFW system supply lines and sources ( e.g.,
condensate storage tanks) was due to relatively active failure mechanisms ( e.g., corrosion). Although these
.components have relatively low pressures, failure of 4.1 these components would disable the entire AFW sys-tem, and thus contribute significantly to core damage.
Relatively high rankings were estimated for the pipe segments of the LPI-Accumulator system extending from the accumulator discharge headers to the RCS isolation valves. The high rankings of these lines are due to their important safety functions in providing coolant to the RCS following an accident. Because the residual heat removal (RHR) system is connected to these lines, ruptures of these lines could prevent cooling water from being supplied to the RCS loops during the plant shutdown or cooldown.
Pipe segments of the LPI system extending from the inside containment isolation valves to the RCS cold-and hot-leg injection headers were also identified to be high risk-important components. The high rankings are due to relatively high stresses, potential for overpres-surization of these lines, and the important functions of these lines in providing coolant to the RCS following a large LOCA. A high risk importance is also noted for the LPI supply lines and water sources ( e.g., refueling water storage tank and containment sump). Failures of these lines result in a total loss of the LPI system. The LPI pump discharge lines up to the containment isola-tion valves follow in risk importance due to either lower stresses and/ or lower core damage consequences result-ing from component ruptures. Depending on the break location, ruptures of these lines result in an interfacing system LOCA outside the containment.
4.1.2 Medium-Risk Importance Components The control rod drive mechanisms (CRDMs) and in-strument lines of the reactor vessel are next in ranking.
These components have much lower estimated conse-quences of failures compared with other RPV compo-nents. The failure consequences were estimated in the worst case to be equivalent to a large LOCA. Equal importance was estimated for the pipe segments of the AFW system extending from the containment isolation valves to the steam generator isolation valves, and also the pipe segments extending from steam supply lines to the AFW pump turbine drive. The importance of these lines is due to a combination of high stress and high system unavailability resulting from a line rupture.
NUREG /CR-6181
4.0 Discussions As shown in Table 3.1, the next most important seg-ments are the main RCS piping from the cold-leg loop stop valves to the pressure vessel, the pressurizer spray piping, and the pipe segments from the pressure vessel to the hot-leg loop stop valves. Failure of any of these lines results in a large LOCA which cannot be isolated by the loop stop values, as is the case with other seg-ments of the main RCS piping. High estimated con-sequences resulting from lines being connected to the RCS loop (e.g., safety injection, RHR lines, etc.) in-crease the importance of these particular pipe segments within the reactor coolant loop. The importance of the pressurizer spray line results from a relatively high-estimated failure probability due to thermal stresses and the key function of this line in controlling the desired primary system pressure. Failure of the spray line could result in LOCAs in Loops A and C, in addition to the loss of the pressurizer function.
Of equal importance in Table 3.1 are the pump suction lines of the LPI system, the AFW turbine-driven pump discharge lines, the Unit 2 AFW pump cross-connected line, and the pipe segments extending from the AFW isolation valves to the steam generators. Ruptures of the LPI pump suction lines would prevent the borated water from being supplied to the RCS when needed.
Similarly, a break in the AFW turbine-driven pump discharge line could prevent sufficient cooling water from being supplied to the steam generators. Unless appropriate recovery actions are taken, a complete loss of water supply to the steam generators may result, contributing significantly to core damage. The impor-tance of the Unit 2 AFW pump cross-connected line is due to its key function in providing cooling to the steam generators in the case that Unit 1 AFW is lost. This cross-connected line is used for mitigating other initi-ating events as well (e.g., station blackout). Finally, failures of the pipe segments extending from the AFW isolation valves to the steam generators would result in steam generator blowdown through the break ( similar to a main steam line break) and a loss of secondary cooling.
4.1.3 Low-Risk Importance Components Lower importances are noted for the RPV nozzle to vessel welds, vessel studs, the RPV upper closure head and flanges, and the RPV nozzle forging (inlet and outlet). The failure consequences of these components NUREG/CR-6181
.,I.,,
4.2 were estimated in the worst case to be equivalent to a large LOCA. The AFW motor-driven pump suction and discharge lines, the turbine-driven pump suction line, the pipe segments from the emergency makeup system, and the fire main were estimated to have rela-tively low rankings. This is due mainly to the low esti-mated consequences for failures of these components.
Of the 37 components ranked in Table 3.1, the follow-ing were identified to have the lowest importance: the pressurizer relief and safety lines, pressurizer surge line, pipe segments extending from the hot-leg loop stop valve to the steam generator, pipe segments extending from the coolant pump to the cold-leg loop stop valve of the RCS, and the accumulator discharge and suction lines. These low rankings are due to low rupture prob-ability estimates and/or low core damage consequence estimates for these components. Additionally, failures of the lines within the RCS loops, at most, will result in a large LOCA which can be isolated by the loop stop values.
The cumulative risk contribution for all 37 components
- of the four systems ( as shown in Figure 3.6) is about 2.lE-06 per plant year. Significant contributions to risk come only from failures of approximately the first 10 components. The welds of the beltline region of the reactor pressure vessel dominate the risk, accounting for almost 75% of the core damage frequency due to component failures. The beltline welds are followed in importance by the beltline plate material, which accounts for another 5%. The welds in the upper and lower reactor heads account for another 6%; the single AFW condensate storage tank and supply line contrib-ute 3%; and various welds in the LPI system contribute another 10%). This adds up to more than 99% of the.
- total core damage frequency risk associated with com-ponent ruptures for the four system analyzed. The system level rankings derived from component contribu-tions to core damage are the following: 1) RPV, 2)
LPI, 3) AFW, and 4) RCS. These results agree with those obtained from the earlier PNL system-level study (Vo et al. 1989).
Table 4.1 presents the Surry-1 plant-specific ASME classifications and required ISI examinations for each piping section or component of Table 3.1. Table 4.1 shows that ASME classifications and ISi requirements are in general agreement with the importance rankings based on cor~ damage frequency. In particular, the
RPV component making the greatest contribution to the core damage frequency has the most stringent in-spection requirements. However, recommendations for setting inspection requirements based solely on Table 4.1 should be made cautiously because additional plant systems will be considered in future PNL work.
4.2 Development of Target Risk and Rupture Probability Values A philosophy and approach for selecting target risk values and target rupture probabilities has been recom-mended by the ASME Research Task Force on Risk-Based Inspection. The philosophy is that the inspection should ensure that the risk of core damage resulting from pressure boundary and structural failures should be a small fraction of the total core damage risk esti-mated in the plant specific PRAs. The ranking process described in this report can be used to set priorities for inspection but does not provide criteria for determining the degree of inspection. For this purpose, the risk due to pressure boundary and structural failures is herein eferred to as the "target risk," and 5% of the total RA estimated risk from internal events has been recommended as an appropriate value. It is further recommended that this target risk be apportioned un-equally among the risk-important components by con-sidering the estimated risk associated with rupture of each component (Gore, et al. 1991). Using the results of conditional probability of core damage given compo-nent failures, then, the target rupture probabilities for components can be estimated.
The core damage frequency estimated in state-of-the-art PRAs for modern facilities is about 5.0E-05 per plant year. Using this number, an appropriate target risk value to be distributed among the components would be 2.5E-06 per plant year (i.e., 5% of the total PRA risk).
4.3 4.0 Discussions Examination of Table 3.1 shows that the total risk for the four systems analyzed is about 2.lE-06 per plant
- year, which is rather close to the total recommended target risk of 2.5E-06 per plant year. Thus, the total
- target risk in this case is essentially the same as the total estimated risk due to ruptures. However, analysis of the remaining systems at Surry-1 may increase this total risk until it may equal or exceed the total recom-mended target risk.
The recommendation that the target risk be appor-tioned on the basis of the component's estimated risk, in this case, means that the target risk values can be set equal to the component's estimated risk. This, in turn, means that the target rupture probability for each com-ponent can be the same as the rupture probability estimated by the expert judgment elicitation panel (since consequences are fixed by the PRA analysis given that a rupture occurs). Table 3.3 of Section 3.0 shows the rupture probabilities for components of Table 3.1.
For example, the beltline region of the Surry-1 RPV contains five welds (four longitudinal welds and one circumferential weld). The per-weld rupture probability was estimated as 3.16E-07, and the estimated target risk for all five welds was 1.58E-06 (Table 3.1). Based on the above discussions, the value of 3.17E-07 can be used as a desired target rupture probability for each weld within the RPV beltline region.
Following this approach, the objective of ISi is to pro-vide confidence that the failure probabilities 'do remain
. at or below the values estimated by the expert elicita-tion. In those cases where the total risk for all com-ponents exceeds target values ( e.g., 2.5E-06 per plant year), and an additional objective of ISi should be the reduction of failure probabilities for selected compo-nents.
NUREG /CR-6181
4.0 Discussions RPV RPV RPV AFW RPV LPI-A LPI LPI LPI LPI LPI RPV RPV AFW AFW RCS LPI RCS RCS Table 4.1. Component Importance Compared with ASME BPVC Section XI Classifications and ISi Requirements for Selected Systems at Surry-1 (a)
ASME BPVC System System-Component Category Examination Beltline Welds B-A Volumetric Beltline Plate B-A Volumetric Lower /Bottom Shell B-A Volumetric CST, Supply Line D-B Visual Circumferential Flange to Nozzle Course, Upper B-A Volumetric Shell, Outside Beltline Welds Pipe Segment Between Accumulator Discharge B-I Volumetric and Header and RCS Isolation Valves Surface Pipe Segment Between Containment Isolation Valve B-J, C-F-1 Volumetric and (inside) and Cold-Leg Injection Surface Pipe Segment Between Containment Isolation Valve B-J, C-F-1 Volumetric and (inside) and Hot-Leg Injection Surface LPI Sources (RWST, Sump), Supply Line D-C Visual Pipe Segment Between Pump Discharge and Con-C-F-1 Volumetric and tainment Isolation Valve Visual Pipe Segment Between Containment Isola~ion Valves C-F-1 Volumetric and Visual CRDMs B-E Visual Instrument Lines B-J Exempt Pipe Segment Between Containment Isolation and C-F-1 Volumetric and SG Isolation Valves Surface Main Steam to AFW Pump Turbine Drive C-F-1, Volumetric, Pipe Segment Between Loop Stop Valve and RPV B-J Volumetric (Cold Leg)
LPI Pump Suction Line C-F-1 Volumetric and Surface Pressurizer Spray Line B-J Volumetric Pipe Segment Between RPV and Loop Stop Valve B-J Volumetric (Hot Leg)
NUREG /CR-6181 4.4
4.0 Discussions Table 4.1 (cont'd)
ASME BPVC System System-Component Category Examination AFW AFW TD Pump Discharge Line D-B Visual AFW Pipe Segment from Unit 2 AFW Pumps C-F-1, Volumetric AFW AFW Isolation Valve to SG C-F-1 Volumetric and Surface RPV Nozzle to Vessel Welds B-D Volumetric RPV Vessel Studs B-G-1
- Volumetric, Surface, and Vi-sual AFW AFW MDP Suction Line D-B Visual AFW AFW MDP Discharge Line D-B Visual RPV Upper, Closure Head, Flange B-A Volumetric and Surface RPV Nozzle Forging Inlet/Outlet B-D Volumetric AFW AFW TDP Suction Line D-B Visual AFW Pipe Segment from Emergency Makeup System, Fire D-B Visual Main RCS Pressurizer Relief/Safety Line B-J Volumetric RCS Pressurizer Surge Line B-J Volumetric LPI-A Accumulator Discharge Line B-J, C-F-1 Volumetric and Surface RCS Pipe Segment Between SG and RCP B-J Volumetric RCS Pipe Segment Between Loop Stop Valve and SG B-J Volumetric (Hot Leg)
RCS Pipe Segment Between RCP and Loop Stop Valve B-J Volumetric (Cold Leg)
LPI-A Accumulator, Su.ction Line C-F-1 Volumetric and Surface
( a)
Based on Surry-1 plant-specific system classifications.
4.5 NUREG/CR-6181
5.0 Summary* and Conclusions A method for planning inspections has been developed and has been applied in a pilot study to identify and.
prioritize the most risk-important systems and compo-nents at Surry-1. In the pilot application, the method used component failure probabilities estimated from the expert judgment elicitation conducted by PNL, results from PNL's Surry-1 system prioritization, and system fault trees reported in the Surry-1 PRA to prioritize to the high-priority components for inspection.
As shown in Table 3.1, contributions of component failures to core damage frequency range widely from about l.6E-06 to 1.6E-14 per plant year. The cumula-tive risk contribution ( as shown in Figure 3.5) is about 2.lE-06 per plant year. This estimate is about 5% of the total Surry-1 PRA risk. The total estimated risk is dominated by failures of the reactor pressure compon-ents (86% ). This risk is followed by the LPI system components (10% ), and then other components within the AFW and RCS (4%). The results provide a guide to establish improved inspection priorities for nuclear power plant components.
sitivity analyses addressed uncertainties on parame-values and modeling assumptions. The sensitivity of mponent rankings to upper-and lower-bounding values of estimated rupture probabilities was estab-lished. As shown in Tables 3.1 and 3.2, the results indicated no significant changes in component rankings.
Additional sensitivity analyses addressed contributions to core damage frequency due to indirect effects of component failures. The results indicate that the over-all contribution to core damage frequency from the indirect effects was negligible. Sensitivity and uncer-tainty analyses regarding potential floodings within the plant due to pipe ruptures have not yet been addressed.
Flooding from some rupture locations could* disable safety-related equipment, thereby contributing signifi-
. cantly to core damage. This important issue will be addressed in future work.
Risk-based priorities were compared with the current Surry-1 plant-specific ASME classifications and required ISi examinations. The ASME classifications and ISi requirements are in general qualitative agreement with risk-based rankings based on core damage frequency.
The components making the greatest contributions to the core damage frequency currently have the most stringent inspection requirements. However, final con-clusions for setting inspection requirements should await further pilot studies.
An approach for determining target risk and target rupture probability values has been proposed and sue~
cessfully pilot tested for components within the selected systems at Surry-1. It is recognized that in some cases the estimated target values of rupture probabilities may be difficult to achieve, therefore, further studies are n~ded to determine whether this approach is generally
- appropriate.
5.1 The analysis for the Surry-1 plant will be completed by developing the risk importance of components in the remaining systems ( e.g., high-pressure injection, service water, and balance of the plant). Similar analyses will be performed for other PWRs and boiling-water reac-
. tors (BWRs), and generic trends in component impor-tances will be established. Once the high-priority com-ponents have been identified, recommended inspection programs (method, frequency and extent) will be devel-oped. Probabilistic structural mechanics and decision analyses will be applied to establish inspection strategies that will ensure that component failure rates are main-tained at acceptable levels and in a cost-effective man-ner.
NUREG /CR-6181
6.0 References
~
Resea<ch Task Force on llisk-Based Inspection Guidelines. 1991. Risk-Based Inspection - Development of Guidelines, Volume 1 General Document. CRTD-Vol, 20-1, American Society of Mechanical Engineers Center for Research and Technology Development.
Bertucio, R. C., and J. A. Julius. 1990. Analysis of Core Damage Frequency: Surry, Unit 1 Internal Events.
NUREG/CR-4550, Sandia National Laboratories, Albu-querque, New Mexico.
Gore, *B. F., T. V. Vo, and K. R. Balkey. 1991. "Status of ASME Rick-Based Inspection Guidelines Develop-ment for Nuclear Power Plants." To be presented at the 1991 ASME Winter Annual Meeting, Atlanta, Georgia.
Heinecke, C. C., V. A. Perone, M. Weaver, and C. N:
Wright. 1987. Surry Units 1 and 2 Reactor Vessel Flu-ence and RTPTS Electrons. WCAP-11015, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania.
Meyer, M.A., and J.M. Booker. 1989. Eliciting and Analyzing &pert Judgment. NUREG/CR-5424, Los os National Laboratory, Los Alamos, New Mexico.
ussel, K. D., et al. 1987. Integrated Reliability and Risk Analysis System (IRRAS). NUREG/CR-4844, Idaho National Engineering Laboratory, Idaho Falls, Idaho.
U.S. Nuclear Regulatory Commission (NRC). 1981.
Standard Review Plan 3.6.2 Detennination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping. NUREG-0800, Rev. 1, U.S. Nuclear Regulatory Commission, Wasliington, D.C.
6.1 U.S. Nuclear Regulatory Commission (NRC). 1989.
Severe Accident Risks: An Assessment for Five U.S.
Nuclear Power Plants. NUREG-1150, Summary Report, Second Draft For Peer Review, U.S. Nucl~ar Regulato-ry Commission, Washington, D.C.
Vo, T. V., B, F. Gore, L. J. Eschbach, and F. A.
Simonen. 1989. "Probabilistic Risk Assessment-Based Guidance for Piping Inservice Inspection." Nuclear Technology, Volume 88 (1), American Nuclear Society, La Grange Park, Illinois.
Vo, T: V., B. W. Smith, F. A. Simonen, and S. R. Doc-tor. 1990. "Development of Generic Inservice Inspec-
. tion Guidance for Pressure Boundary Systems. Nucle-ar Technology, Volume 92 (3), American Nuclear Soci-ety, La Grange Park, Illinois.
Vo, T. V., P. G. Heasler, S. R. Doctor, F. A. Simonen, and B. F. Gore. 1991. "Estimate of Component Rup-ture Probabilities. Expert Judgment Elicitation." Nu-clear Technology, Volume 94 (1), American Nuclear Society, La Grange Park, Illinois.
Wheeler, T. A., S. C. Hora, W.R. Cramond, and S. D.
Unwin. 1989. Analysis of Core Damage Frequency from Internal Events: &pert Judgment Elicitation.
NUREG/CR-4550, Volume 2, Sandia National Labora-tories, Albuquerque, New Mexico.
Appendix A Sample of Component Importance Calculations NUREG /CR-6181
Appendix A Sample of Component Importance Calculations This appendix shows the component risk importance calculations for the reactor pressure vessel. Similar calculations were performed for the other systems (reactor coolant, low-pressure injection, and auxiliary feedwater) that are addressed in this report.
PROGRAM:
SYSTEM:
NOE-IMPROVED IS!
REACTOR PRESSURE VESSEL FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET COMPONENT:
FLANGE TO NOZZLE/UPPER SHELL WELD
( 1)
(2)
(3)
(4)
(5)
Failure Component Probabi lit) a)
Core Damffi (location)
(component size)
Failure Effect Recovery Action Frequency Weld 1 - circumfer-( 157" ID Core damage was No recovery 2.00E-08 ent i al flange to 9" thick) assumed nozzle course 3.16E-08 p cm/s = 1. 0 2.00E-08 1.00E-08 P/Pf = 1. 0 CDP= 1. 0 Weld 2 - circumfer-(157" ID Core damage was No recovery 3.16E-08 ential upper shell 9" thick) assumed to weld, intermedi-ate shell
- 1. OE-07 p cm/s = 1. 0 3.16E-08
- 1. OE-08 P/Pf = 1.0 CDP = 1.0 SHEET 1 OF 13 (6)
(7)
ImportfBJe Index Remarks 4
- Low fluence and low embrittlement
- The RPV was assumed to fail given a break at this location 3
- Lower fluence and lower embrittlement
- Cu content= 0.11%
- Ni content= 0.7%?
(a)
(bl (c)
Component failure probability obtained.from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
p
= Pf*~ p I
- P I Pf*R I cm jt cm s1 s1 1
(d)
Based on "Median Values" of failure probabilities.
NOTE:
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
~
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PROGRAM:
SYSTEM:
COMPONENT:
( 1)
Component (location)
Weld 3 - lower shell NDE-IMPROVED !SI REACTOR PRESSURE VESSEL LOWER SHELL/BOTTOM HEAD (2)
Failure Probability(a)
(component size)
(157" ID 9" thick)
- 1. OE-06 3.16E-07
- 1. OE-07 Weld 4 - bottom (157" ID head 5.4" thick)
- 1. OE-07 3.16E-08 5.2E-09 FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 2 DF 13 (3)
(4)
(5)
(6)
(7)
Fai 1 ure Effect Recovery Action Core Damffi Frequency ImportfBJe Index Remarks Core damage was No recovery 3.16E-07 1
- High fluence and high assumed embrittlement.
High ther-mal stress for LOCA and p
= 1. 0 PTSs.
cm/s
- Critical weld based on P/Pf = 1.0 Surry PTS study (WCAP-11015, 1987).
CDP = 1.0
- Cu content= 0.11%
- Ni content= 0.7%?
Core damage was No recovery 3.16E-08 3
- Lower fluence and lower assumed embrittlement
- Cu content= 0.11%
pcm/s = 1. 0
- Ni content= 0.7%?
P/Pf = 1.0 CDP = 1. 0 (a)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl (cl (d)
NOTE:
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
pcm= Pf*~ pcm/s1*PsJ Pt'*R1I Based on "Median Values" of failure probabilities.
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
PROGRAM:
SYSTEM:
NOE-IMPROVED ISi REACTOR PRESSURE VESSEL FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET COMPONENT:
BOTTOM HEAD/INTERMEDIATE SHELL
( 1)
(2)
(3)
(4)
(5)
Fai 1 ure Component Probability(a)
Core Dam'.Y:m (location)
(component size)
Failure Effect Recovery Action Frequency Weld 5 - bottom
( 157" ID Core damage was No recovery 3.16E-08 head 5.4" thick) assumed
- 1. E-07 p
- 1. 0 3.16E-08 cm/s -
Core damage was No recovery 3.16E-07 nal weld intermedi-assumed ate shell 5E-07 3.16E-07 p cm/s = 1. 0
- l. lE-07 P/Pf = 1.0 CDP = 1.0 SHEET-3 OF 13 (6)
(7)
ImportfBJe Index Remarks 3
- Low fluence and low thermal stress.
1
- High fluence and embrit-tlement at lower end of this weld.
High thermal stress for LOCA and PTS.
- Cu content= 0.11%
- Ni content= 0.7%?
(a)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
( b J Importance measure or rank relative to other components of the RPV, based on "Medi an Values" of failure probabilities.
(c)
Pcm= P.e*~ Pcm/s, *Ps/ P.f*R1I (d)
Based on "Median Values" of failure probabilities.
NOTE:
Q - never used on these tables.
CDP - core damage probability given the break.
EOP - emergency operating procedure.
. LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
~
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PROGRAM:
SYSTEM:
COMPONENT:
NOE-IMPROVED ISI REACTOR PRESSURE VESSEL INTERMEDIATE SHELL (1)
(2)
Failure Component Probability(a)
. (location)
(component size)
Weld 7 - longitudi-(100" length) nal weld interme-di ate she 11 5.0E-07 3.16E-07
- 1. OE-06 3.16E-07
- l. lE-07 FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 4 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Dam'.'{ir)
Freauency ImportfBJe Index Remarks Core damage. was No recovery 3.16E-07 1
- High fluence and embrit-assumed tlement at lower end of this weld.
High thermal p
= 1. 0 stress for LOCA and PTS.
cm/s
- Cu content= 0.11%
P/Pf = 1.0
- Ni content= 0.7%?
CDP = 1. 0 Core damage was No recovery 3.16E-07 1
- High fluence and embrit-assumed tlement at upper end of this weld.
High thermal p cm/s = 1. 0 stress for LOCA and PTS.
- Critical weld based on P/Pf = 1.0 Surry PTS study (WCAP-11015, 1987).
CDP =.1.0
- Cu content= 0.11%
- Ni content= 0.7%?
(a)
Component failure probability obtained from Vo et al. 1991.. The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl (c)
(d)
NOTE:
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
Pcm= Pf*~ Pcm/s1 *Ps/ Pf*R1I Based on "Medi an Values" of fa.il ure probabi 1 it i es.
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
PROGRAM:
SYSTEM:
NOE-IMPROVED !SI REACTOR PRESSURE VESSEL FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET COMPONENT:
INTERMEDIATE SHELL/NOZZLE TO VESSEL WELD (1)
(2)
(3)
(4)
(5)
Failure Component Probabil i ty(a)
Core Damffi (location)
(comoonent size)
Failure Effect Recovery Action Frequency Weld 9 - longitudi-(100" length)
Core damage was No recovery 3.16E-07 nal weld lower assumed shell
L/LOC,A (see re-Follow EOPs
- 1. OOE-11 to-vessel weld, mark)
Loop 1 outlet 3.16E-08 2.0E-08 pcm/s = 5.00E-04
- 1. lE-08 P/Pf = 1.0 CDP= 5.00E-04 SHEET 5 OF 13 (6)
(7)
ImportfBJe Index Remarks 1
- High fluence and embrit-tlement at upper end of this weld.
High thermal stress for LOCA and PTS.
- Critical weld based on Surry PTS study (WCAP-11015, 1987).
- Cu content= 0.11%
- Ni content= 0.7%?
7
- Lower fluence and low embrittlement.
- Break at this location was assumed to be equiva-lent to a large LOCA.
(a)
(b)
(c)
(d)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
Pcm= P.t'*l'f' Pcm/s *P8
/ P.t'*RJ rt 1
1
~
Based on "Median Values" of failure probabilities.
NOTE:
Q - never used on these tables.
CDP - core damage probability given the break.
EOP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
~
~
0
~
O'I I-'
00 I-'
PROGRAM:
SYSTEM:
COMPONENT:
NOE-IMPROVED ISI REACTOR PRESSURE VESSEL NOZZLE TO VESSEL WELDS (1)
(2)
Failure Component Probability(a)
(location)
(component size)
Weld 11 - nozzle-(app. 40" dia.)
to-vessel weld, Loop 1 inlet 3.16E-08 2.0E-08
- 1. OE-08 Weld 12 - nozzle-(app. 40" dia.)
to-vessel welds, Loop 2 outlet 3.16E-08 2.0E-08
- 1. OE-08 FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 6 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Dam'.¥1:n Frequency ImportfBJe Index Remarks L/LOCA was No recovery
- 1. OOE-11 10
- Lower fluence and low assumed given a embrittlement.
break Pcm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 L/LOCA was as-Follow EOPs
- 1. OOE-11 10
- Lower fluence and low sumed given a embrittlement.
break Pcm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 (a)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl (cl (d)
NOTE:
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
p= = Pf*~ pcm/s1 *Ps1f Pf*Ri' Based on "Median Values" of failure probabilities.
Q - never used on these tables.
CDP - core damage probability given the break.
EOP - emergency operating procedure.
LOCA - loss of coolant accident.
- PTS - pressurized thermal shock.
PROGRAM:
SYSTEM:
COMPONENT:
NOE-IMPROVED IS!
REACTOR PRESSURE VESSEL NOZZLE TO VESSEL WELDS
( 1)
(2)
Failure Component Probabi 1 ity (a)
(location)
(component size)
Weld 13 - nozzle-(app. 40" di a.)
to-vessel weld, Loop 2 inlet 3.16E-08 2.0E-08 l.OE-08 Weld 14 - nozzle-(app. 40" dia.)
- to-vessel welds, Loop 3 outlet 3.16E-08 2.0E-08
- l. OE-08 FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 7 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Damffi Frequency ImportfBJe Index Remarks L/LOCA was Foll ow EOPs l.OOE-11 10
- Lower fluence and low assumed given a embrittlement.
break pcm/s = 5.00E-04 P/Pf = l. 0 CDP = 5.00E-04 L/LOCA was as-Fo 11 ow EOPs l.OOE-11 10
- Lower fluence and low sumed given a embrittlement.
break p cm/s = 5.00E-04 P/Pf = 1.0 CDP = 5. OOE-04 (a)
- component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(b)
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c)
Pcm= Pt*~ Pcm/s1 *Ps/ Pt*Ril (d)
Based on "Median Values" of failure probabilities.
NOTE:
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
~
~
t'I1 Cl -
Q
°'
I-'
00 I-'
PROGRAM:
SYSTEM:
NOE-IMPROVED ISI REACTOR PRESSURE VESSEL FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET COMPONENT:
NOZZLE TO VESSEL WELD/NOZZLE FORGINGS (1)
(2)
(3)
(4)
(5)
Fai 1 ure Component Probabi 1 it)al Core Dam'.Y:tr)
(location)
(component size)
Failure Effect Recovery Action Freauency Weld 15 - nozzle-(app. 40" dia.)
L/LOCA was Follow EOPs
- 1. OOE-11 to-vessel weld, assumed given a Loop 3 inlet 3.16E-08 break 2.0E-08
- 1. OE-08 Pcm/s = 5.DOE-04 P/Pf = 1.0 CDP= 5.00E-04 Nozzle Forgings -
(app. 40" dia.)
L/LOCA was as-Foll ow EOPs 1.50E-11 inlets for Loops 1, sumed given a 2, and 3 5.0E-08 break 3.0E-08
- 1. OE-08 p cm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 SHEET 8 OF 13 (6)
(7)
ImportfBJe Index Remarks 10
- Lower fluence and low embrittlement.
10
- High stress at nozzle corner.
Potent i a 1 for high thermal stress from cold fluid injection.
(a)
Component fai 1 ure probabi 1 i ty obtained from Vo et al. 1991.
The three va 1 ues repres*ent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl (c)
(d)
NOTE:
Importance measure or rank relative to other components of the RPV, based on "Median Va 1 ues" of failure probabilities.
Pcm = Pt*~ pcm/s1 *Ps,f Pt*R1I Based on "Median Values" of failure probabilities.
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
PROGRAM:
SYSTEM:
NOE-IMPROVED ISi REACTOR PRESSURE VESSEL FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET COMPONENT:
NOZZLE FDRGINGS/BELTLJNE PLATE (1)
(2)
(3)
(4)
(5)
Failure Component Probability(a)
Core Dam'.¥tr)
(location)
(component size)
Fai 1 ure Effect Recovery Action Frequency Nozzle Forgings -
(app. 40" dia.)
L/LOCA was Follow EOPs
- 1. DDE-11 outlets for Loops assumed given a 1, 2, and 3 5.DE-08 break 2.DE-08
- 1. DE-08 Pcm/s = 5.0DE-04 P/Pf = 1.0 CDP= 5.00E-04 Beltl ine Plate (157" JD Core damage was No recovery
- 1. ODE-07 (base metal) 9" thick) assumed
- 1. DE-07 p
- 1. D
- 1. DE-07 cm/s -
8.DE-08 P/Pf = 1.0 CDP = 1.0 SHEET 9 OF 13 (6)
(7)
ImportfBJe Index Remarks 10
- High stress at nozzle corner.
2
- High fluence, but lower embrittlement rate than welds.
(a)
(b)
(c)
(d)
NOTE:
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
p
= P.)Y' p I *P I P~*R*I cm L jt cm s 1 s1 L
i Based on "Median Values" of failure probabilities.
Q -*never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
~
~
t'I1 0 -
~
°'
I-'
00 I-'
PROGRAM:
SYSTEM:
COMPONENT:
(1)
Component (location)
Vessel Shell NOE-IMPROVED ISI REACTOR PRESSURE VESSEL VESSEL SHELL/UPPER HEAD (2)
Fai 1 ure Probability(a)
(component size)
(157" OD outside beltline 9" thick)
(base metal) 3.16E-08
- 1. OE-08
- 1. OE-08 Upper Head (base (157" ID metal) 6" thick) 5.DE-08 3.16E-08
- 1. OE-08 FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 10 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Damffi Frequency ImportfBJe Index Remarks Core damage was as-No recovery 1.0DE-08 5
- Lower fluence and less sumed given a break at embrittlement.
this location p cm/s = 1. O P/Pf = 1.0 CDP = 1. 0 L/LOCA was assumed No recovery
- 1. 58E-11 9
- Low fluence and low given a break at this embrittlement.
location p cm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 (a)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl (c)
(d)
NOTE:
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
p
= Pf*IY" p I *P I Pf*R1I cm rt cm s 1 s 1 Based on "Median Values" of failure probabilities.
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
PROGRAM:
SYSTEM:
COMPONENT:
( 1)
Component (location)
NOE-IMPROVED !SI REACTOR PRESSURE VESSEL LOWER HEAD/VESSEL FLANGE (2)
Failure Probability(a)
(component size)
Lower Head (base (157" OD metal) 5.4" thick) 5.DE-08
- 1. OE-08
- 1. OE-08 Vessel Flange
( 149" ID x 184" OD) 5.0E-08 8.0E-08
- 1. DE-08 FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 11 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Damffi Freauencv ImporttBre Index Remarks Core damage was No recovery
- 1. ODE-08 5
- Lower fluence and low assumed embrittlement.
p cm/s = 1. a P/Pf = 1.0 CDP = 1. 0 L/LOCA was as-No recovery l.OOE-11 10
- Low fl uence.
Stress sumed given the concentr~tion from closure break studs.
- Potential surface imper-p cm/s = 5.00E-04 fection.
However, it has been eliminated; there-P/Pf = 1.0 fore, creating more reli-able flange surface.
CDP= 5.00E-04 (a)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, an*d 25% and 75% quartiles of uncertainty distribution.
(b)
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c)
Pcm= Pt*~ pcm/s1*Ps/ Pt*Ril (d)
Based on "Median Values" of failure probabilities.
NOTE:
Q - never used on these tables.
COP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
j
~
~
tI1 0
("")
~
0\\
00 PROGRAM:
SYSTEM:
COMPONENT:
(1)
Component (location)
Closure Head Flange Reactor Vessel NOE-IMPROVED !SI REACTOR PRESSURE VESSEL VESSEL FLANGE/STUDS (2)
Failure Probability(a)
(component size)
(149" ID x 184" OD) 5.0E-08 2.0E-08
- 1. OE-08 (app. 6" dia.
Studs (58 studs) each)
- 1. OE-06
- 1. OE-07 3.16E-08 FAILURE MODES AND EFFE ALYSIS WORKSHEET SHEET 12 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Damffi Frequency ImportfBJe Index Remarks L/LOCA was as-No recovery l.OOE-11 10
- Low fl uence.
Stress concen-sumed given a tration from closure studs.
break at this
- Potential leakage at RPV location flanged leak off line (1" line) due to 0-ring or valves p cm/s = 5.00E-04 (upstream) leakage.
P/Pf = 1.0 CDP= 5.00E-04 L/LOCA was as-No recovery 5.00E-11 8
- Multiple failure of vessel sumed given the studs must be postulated for break (see re-LOCA.
marks)
- Some stud nicks -and gauge problems.
p cm/s = 5.00E-04
- Multiple stud failures con-tribute a small fraction of P/Pf = 1.0 core damage frequency (e.g.,
Chi-square distribution to CDP= 5.00E-04 quantify contributions from multiple stud failures).
(a)
Component failure probability obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(b)
(c)
(d)
NOTE:
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
Pcm = Pt*~ Pcm/s1 *Ps1 / Pt*Ril Based on "Median Values" of failure probabilities.
Q - never used on these tables.
CDP - core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
~
0 -
~
I-'
00 I-'
PROGRAM:
SYSTEM:
COMPONENT:
NOE-IMPROVED ISi REACTOR PRESSURE VESSEL CRDNs/INSTRUMENT LINES (1)
(2)
Failure Component Probability(a)
(location)
(component size)
Control Rod Drive (app. 4" dia.)
Mechanism (~5 penetrations of 5.48E-05 upper head)
- 1. OE-05 4.47E-06 Instrument Line (app. 1" di a.)
Penetrations (lower head, 1o'o 8.37E-05 instrument lines)
- 1. DE-05 3.16E-06 Lower Head Skirt FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 13 OF 13 (3)
(4)
(5)
(6)
(7)
Failure Effect Recovery Action Core Damffi Frequency ImportfBJe Index Remarks L/LOCA was as-No recovery 5.DOE-09 6
- Contribution from multiple sumed given a CROM failures were insignifi-break at (see cant (e.g., using Chi-square remarks) distribution to estimate the failure effect due to multiple p cm/s = 5.00E-04 CROM fail u res).
P/Pf = 1. D CDP= 5.DOE-04 L/LOCA was as-No recovery 5.DDE-09 6
- Cracking and repairs have sumed been reported at some plants.
- Seal table (thimble tube) p cm/s = 5.0DE-04 leakage due to seal fitting failures or out of adjustment.
P/Pf = 1.0
- Contribution to CDP from multiple IL failures was esti-CDP= 5.0DE-04 mated to be insignificant.
- Laying on concrete support.
(a)
(bl (c)
(d)
NOTE:
Component failure probabflity obtained from Vo et al. 1991.
The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
p
= Pf*IY' p I *P I Pf*R1I cm rt cm s 1 s1 Based on "Median Values" of failure probabilities.
Q - never used on these tables.
CDP~ core damage probability given the break.
EDP - emergency operating procedure.
LOCA - loss of coolant accident.
PTS - pressurized thermal shock.
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,1 t NRC FORM 335 12-891 U.S. NUCLEAR REGULATORY COMMISSION I 1. REPORT NUMBER NRCM 1102, 3201, 3202
- 2. TITLE ANO SUBTITLE BIBLIOGRAPHIC DATA SHEET (SH instructions on rht1 rt1VtlrStlJ A Pilot Application of Risk-Based Methods to Establish Inservice Inspection Priorities for Nuclear Components at Surry Unit I Nuclear Power Station
- - 5. AUTHOR(Sl T. Vo, 8. Gore, F. Simonen, S. Doctor IAaitMd t,y NRC. Add Vol ** $uclo., RIW.,
,.__'" Nummn, If.. v.l NUREG/CR-6181 PNL-9020
- 3.
CATE REPORT PUBLISHED MONTH YEAR AUCJUSt 1994
- 6. TYPE OF REPORT Technical
- 7. PERIOD COVERED /lnctu:uw, D*r~11 10/89-1/92
- 8. PERFORMING ORGANIZATION - NAME ANO ADDRESS /If NRC.pnwiM Dwillian, Offi,,.arR.-. v.s. NuclHr RllflUt.,ro,y Camrniuian, andmailingMJdrr,u;if canrr..:ror,pm*i<l9 n-* Mtd,,,.;1;115 Mlriffu..J Pacific Northwest Laboratory Richland, WA 99352
- 9. SPONSOR ING ORGANIZATION - NAME ANO AOC RESS /If NRC, ry,,. *-s.m. u *bo.. "; ii canr-ror. pnwm NRC Dillirion. Offic* or R.,,ion, v.s. Nuc1.., RllflUl*rory CammiAion,
.,,,,,.,.;Jing __
Divi~ion of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 -OOOf
- 10. SUPPLEMENTARY NOTES
- 11. ABSTRACT 1200 wot'b or-I As part of the NOE Reliability Program sponsored by-the NRC, PNL is developing a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components.
This method uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis techniques to identify and prioritize th,e most risk-important systems and components for inspection.
The Surry Nuclear Power Station Unit I was selected for pilot applications of this method.
The specific systems addressed in this report are the reactor pressure vessel, reactor coolant, low-pressure injection, and auxiliary feedwater.
The results provide a risk-based ra~king of components within these systems and relate the target risk to target failure probability values for individual components.
These results will be used to guide the development of improved inspection plans for nuclear power plants.
To develop inspection plans, the acceptable level of risk from structural failure for important systems and components will be apportioned as a small fraction of the total PRA-estimated risk for core damage.
This process will determine acceptable target risk and target failure probability values for individual components.
Inspection requirements will be set at levels to assure that acceptable failure probabilities are maintained.
- 12. KEY WOROS/OESCR!PTORS ILi.r IOOldsarpt,,... rn*r will-i,rtWN*n:h,,nin lac*ting rll*flltlOrt.J nondestructive evaluation, probabilistic risk assessment, ASME Code, inservice inspection, welds, piping systems, Inspection Importance, Birnbaum Importance, pressure boundary systems, risk-based, Surry-1 NRC FORM 33S 12-l:19)
- 13. AVAILABILITY STATEMl:NT Unlimited
- 14. SECURITY CLASSIFICATION (This Page/
Unclassified (This Report/
Unclassified
- 15. NUMBER OF PAGES
- 16. PRIC.E