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) | |
Text
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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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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
~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
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Abstract As part of the Nondestructive Evaluation Reliability provide a risk-based ranking of components within Program sponsored by the U.S. Nuclear Regulatory these systems and relate the target risk to target failure Commission, the Pacific Northwest Laboratory is dev- probability values for individual components. These eloping a method that uses risk-based approaches to results will be used to guide the development of im-establish inservice inspection plans for nuclear power proved inspection plans for nuclear power plants. To plant components. This method uses probabilistic risk develop inspection plans, the acceptable level of risk assessment (PRA) results and Failure Modes and Ef-
- from structural failure for important systems and com-fects Analysis (FEMA) techniques to identify and pri- ponents will be apportioned as a small fraction (i.e.,
oritize the most risk-important systems and components 5%) of the total PRA-estimated risk for core damage.
for inspection. The Surry Nuclear Power Station Unit 1 This process will deteI1nine target (acceptable) risk and was selected for pilot applications of this method. The target failure probability values for individual compo-specific systems addressed in this report are the reactor nents. Inspection requirements will be set at levels to pressure vessel, the reactor coolant, the low-pressure assure that acceptable failure probabilities are main-injection, and the auxiliary feedwater. The results 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
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t I 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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I I Executive Summary As part of the Evaluation and Improvement of Nonde- Table S.1 shows the risk importance parameters for structive Evaluation (NDE) Reliability for Inservice Surry-1 components. Included in the table are the Inspection of Light Water Reactors Program sponsored estimated rupture probabilities for the components of by the U.S. Nuclear Regulatory Commission (NRC), the systems analyzed.
the Pacific Northwest Laboratory (PNL) has developed and applied a method using risk-based techniques to On the basis of core damage frequency, contributions of establish inservice inspection (ISi) plans for nuclear component failure to core damage frequency range power plant components. As described in this report, from about 1.6E-14 to 1.58E-06 per plant year. The the method uses probabilities of component failures cumulative risk contribution for all of the components (estimated by using an expert judgment elicitation pro- considered was estimated to be about 2.lE-06 per plant cess) and plant-specific probabilistic risk assessment year. Figure S.3 shows the results of cumulative risk (PRA) results in conjunction with the Failure Modes contribution for Surry-1 components within the systems and Effects Analysis (FMEA) technique to establish ISi analyzed. This estimate is about 5% of the total Surry priorities for systems and components. Included in this PRA risk. The total estimated risk is dominated by report is an approach for determining target risk and failures of the reactor pressure vessel components (86%
target rupture probability values for nuclear power of the total PRA risk). This risk is followed by the low-plant components. pressure injection system components (10%), and then other various components within the auxiliary feedwater The Surry Nuclear Power Station Unit 1 (Surry-1) was and reactor coolant systems (4%). The results provide selected for demonstrating the risk-based methodology. a guide to establish improved inspection priorities for The specific systems addressed in this report are the nuclear power plant components.
reactor pressure vessel, reactor coolant, low-pressure injection including accumulators, and the auxiliary feed- To address uncertainties in the numerical results of the ater systems. The FMEAs were initially formulated study, sensitivity analyses were performed. Based on sing plant system drawings and other plant-specific uncertainties in estimated probabilities of component information. The Standard Review Plan information rupture probabilities, the sensitivity analyses results developed by the NRC was used in determining the indicated no significant changes in component risk effects of system failures. To ensure that the plant rankings (as shown in Figure S.1). Sensitivity analyses models were as realistic as possible, visits at the Surry-1 were also performed to determine overall core damage plant were conducted for plant system walkdowns and frequency due to component failures by indirect effects discussions were held with plant operational and techni- (pipe whip, jet impingement effects, etc.). The results cal staff. Participation of Virginia Electric Power Com- indicate that contributions to the overall core damage pany staff was an essential part of the pilot study. frequency from the indirect effects were negligible.
Because of similarities in objectives, the PNL program Included in the report is a comparison of the risk-based is coordinated with the American Society of Mechanical inspection priorities suggested by this study with the Engineers (ASME) Research Task Force on Risk- current Surry-1 plant ISI practices. ASME classifica-Based Inspection Guidelines. This task force has made tions and ISI requirements are generally in quantitative general recommendations on the application of risk- agreement with the risk-based rankings based on core based methods to inservice inspection and will make damage frequency. The components making the great-specific proposals to ASME for improved codes and est contribution to the core damage frequency have the standards. Results of PNL studies are being made most stringent ASME inspection requirements (i.e.,
available to the ASME group to demonstrate and vali- both volumetric and surface examinations).
date the usefulness of the risk-based concepts.
The analysis for the Surry-1 plant will be completed by The results of the risk-based component prioritization developing the risk importance of components in the are summarized in Figure S.1. For the purpose of remaining systems (e.g., high-pressure injection, service comparison, the components were also ranked on the water, and balance of plant). Similar plant-specific basis of a conditional probability of core damage given analyses will be performed for other pressurized-water a component failure, as shown by Figure S.2. reactors and for boiling-water reactors. Generic trends Vil NUREG /CR-6181
A' Executive Summary in component importance will be established from these plant-specific evaluations. Once the high-priority com-ponents have been identified, recommended inspection 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-programs (method, frequency, and extent) will be devel- age. This process will determine target (acceptable) oped. Probabilistic structural mechanics will be applied risk and target failure probability values for individual to establish inspection strategies that will ensure the components. Inspection requirements will be set at component failure probabilities are maintained at ac- levels to assure that acceptable failure probabilities are ceptable levels. maintained.
NUREG /CR-6181 Vlll *
.. 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
Executive Summary Table S.1 (cont'd)
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 25 and 75 { ::C Median Quartiles 1 Value 10-11
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- - LL
- I
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- - ca l:: E Cl)
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~- 8c 10-11 I!0 0
10-12 1 2
RPV-Beltllne Region Welds RPV*Bellllne P&ate 11 LPI-Plpe Segment Between Containment Isolation Valves 22 23 AFW-lsolatlon Valve to SG RPV-Nozzle to Vessel Welds 3 RPV-Lower/Bottom Shell 12 APV*CADMS 24 RPV-Vessel Studs 2J}J 4 AFW-CST, Supply Line 13 APV*lnstrument Lines 25 AFW*AFW MOP Suction Line 5 RPV*Clr. Flange to Nozzle Course, 14 AFW*Plpe Segment Between Contalnmen1 26 AFW-AFW MOP Discharge Une Upper Shell, Outside BeltUne Welds Isolation & SG Isolation Valves ~ APV*Upper, Closure Head Range LPl*A-PJpe Segment BalwQ8n Acc. 28 RPV-Nozzle Forging Outlet 10-13 Discharge Header & RCS Isolation Valves 15 16 AFW-Maln Steam to AFW Pump Turbine Drive RCS-Pipe Segment Between Loop Stop Vatve 29 AFW*AFW TOP Suction Line 1
7 LPI-Plpe Segmen! Between Containment & APV (Cold Leg) 30 AFW-Pipe Segment from Emergency lsola1lon Valve (Inside) & Cold Lag lnJOC1lon 17 LPl*LPI Pump Suction Line Makeup System, Are Main LPI-Plpe Segment Betw&8n Containment lsol. 18 RCS-Pressurizer Spray Line 31 RCS-Pressurizer ReHef/Satetyu Una Valve (Inside) & Hot Leg Injection 19 RCS-Pipe Segment Between RPV & Loop Stop 32 RCS-Pressurizer Surge line 9 LPJ,Source (RWST,Sump), Supply Line Valve(Hotleg) 33 LPl*A-Aecunwlator Discharge Uno 10*14 10 LPl*Pfpe Segment Between Pump Discharge &
Containment lsol. Velva 20 21 AFW*AFW TD Pump Discharge Una AFW-Pipe Segment from Unit AFW Pumps 34 35 RCS-Pipe Segment Between SG and RCP RCS-Pipe Segment Between Loop 36 ACS-Pipe Segmem Between Stop Valve & SG (Hot leg) RCP & Loop Stop Vatve (Cokt Leg) 37 LPl*A-Accumulator, Suction Una 0 5 10 15 20 25 30 35 Component Identification R9111050 .3 Figure S.1. Risk Contributions of Surry-I Components for the Four Systems Addressed by this Study
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 37 101 10° 10-1 10*2 10-3 10-4
~:
Rani< Syatem Component 10*5 1 RPY*BeHllne Roglon*Weld*
18 20
- RPV-lnatrument Unn RPV*Vnael Studa 2 RPV-BoHllno PlalO 21 RCS.Preasurlzer!R*U*f Safety Un*
3 RPV-1..ower/Bottom Shall 22 ACS-Pra1aurlzer Surge Line 4 RPV..Clr. Flange to Nozzle Course, Upper Shell, Outside Beltllne Wekla 23 RCS-Plpo Soglnonl Bolwoon SQ ond RCP 10-a
- 6
~Saun:H(RWST, Sumpl, Supply Uno LPI-Plpe Segment BetwNn Contalnmant IIOI. Valw (lnalde) I Cold Lag lnjectlon 24 25 RCS-Plpo Segmonl a....... Loop Slop Volw & SQ (lfal l.ogl RCS-Plpo Segmont a....... RCP & Loop Slop Volw (Cold Loal 7 ~Plpo Segmont Be1Wftn Pump DIK!wgo & C...lolnmonl loo, Volvo 28 AFW*Plpe Segment BetwNn* Containment taolallon Yaiva and SQ laollUon V*lvn 8 LPI-A.Plpe Segment BatwNn Acc. Discharge Hnder & RCS laalatlon V.alvH n AFW-lsolallon Valve ID SQ 9 AFW.CST, Suppl~ Uno 28 AFW*AFW TD Pump DIKhorgo u..
10 ~Pipe Segment Bolweon ConlO!nmonl lsol. Volvo (ln11dol 6 Cold Log lnjoclJan 21 AFW*llaln Stum to AFW Pump Turbine Drlve 11 LPf.Plpe Segment Between Contalnment lsol. VatveL 38 AFW*Plpo Sogmonl fn,m Untt AFW Pumpo 12 RCS-Plpo Sogmonl Botweon Loop Slop Volvo & RPV (Cold l.ogl 31 RCS-Pnuurlz<< Spray Uno 10-1 13 14 RPV-Nozzle to Vtnel1 Wekls RCS-Plpo Sogmonl Botween RPV & Loop Slop Volvo (Kol l.ogl 32 33 AFW*AFWTDP Sue11'"1 Une AFW*AFW IIDP Dlaclwgo U..
15 RPY*Upper Cloaure Head Fl.Inge 34
- AFW*AFWIIDP Sucllan Uno 16 ~L.Pt Pump SocUon Uno 35 LP~A-Accumulolor DIKlwgo Uno 17 RPV-No- Forging lnleUOullet 38 AFW*Plpo Segmonl fn,m Ernorgoncy lluoup Sp,..., Flre lloln 18 RPV.CRDIIS 37 LPI-A-Accumulator, Suction UM 10-s 0 5 10 15 20 25 30 35 40 Component Identification 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
- Beltllne Region Welds 15 AFW
- 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 LPI
- A
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
- Pipe Segment Between 23 RPV
- Nozzle to Vessel Welds 1.0 X 10-6 Containment tsol. Valve (Inside) and Hot Leg lnJecUon 24 25 RPV
- Vessel Studs AFW
- 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
I I JI Acknowledgments This work was supported by the U.S. Nuclear Regulato- addressed to the many Virginia Electric Power Compa-ry Commission {NRC) under a Related Service Agree- ny staff for their participation in this work, particularly ment with the U.S. Department of Energy under con- Ms. C. G. Lovett, Mr. R. K. MacManus, Mr. A.
tract DE-AC06-76RLO 1830. The authors wish to McNeil, Mr. D. Rogers, Mr. D. Sommers, and Mr. E.
acknowledge the direction and support provided by Dr. W. Throckmorton. The authors wish to thank Joe Muscara, NRC Program Manager. Dr. L. R. T. T. Taylor, B. W. Smith, and T. W. Bardell for their Abramson from the NRC staff provided guidance to the contributions and review of this work.
elicitation process. Acknowledgments are also xv NUREG/CR-6181
Acronyms AFW Auxiliary Feedwater System ASME American Society of Mechanical Engineers ASTM American Society of Testing and Materials BPVC Boiler and Pressure Vessel Code BWR Boiling Water Re~ctor CRDM Control Rod Driven Mechanisms FMEA Failure Modes and Effects Analysis FSAR Final Safety Analysis Report IRRAS Integrated Reliability and Risk Analysis System ISI Inservice Inspection LOCA Loss of Coolant Accident LPI Low Pressure Injection System LPI/LPR Low Pressure Injection/Recirculation System MDP Motor Driven Pump MOY Motor-Operated Valves NDE Nondestructive Evaluation NRC U.S. Nuclear Regulatory Commission P&ID Piping and Instrumentation Diagram PNL Pacific Northwest Laboratory PRA Probabilistic Risk Assessment PWR Pressurized Water Reactor RCP Reactor Coolant Pump RCS Reactor Coolant System RPV Reactor Pressure Vessel RWST Reactor Water Storage Tank SG Steam Generator TDP Turbine Driven Pump VEPCO Virginia Electric Power Company VIMS Video Information Management System xvii NUREG/CR-6181
Previous Reports in Series
.Doctor, S. R., A. A. Dia7., J. R. Friley, M. S. Good, Heasler, P. G., T. T. Taylor, J. C. Spanner, S. R. Doc-M. S. Greenwood, P. G. Heasler, R. L. Hockey, R. J. tor, and J. D. Deffenbaugh. 1990. Ultrasonic Inspection Kurtz, F. A. Simonen, J.C. Spanner, T. T. Taylor, and Reliability for Intergranular Stress Co"osion Cracks: A T. V. Vo. 1993. Nondestructive Examination (NDE) Round Robin Study of the Effects of Personnet Proce-Reliability for Inservice Inspection of Light Water Reac- dures, Equipment and Crack Characteristics.
tors. NUREG/CR-4469, PNL-5711, Vol. 15. Pacific NUREG/CR-4908. Pacific Northwest Laboratory, Northwest Laboratory, Richland, Washington. Richland, Washington.
Doctor, S. R., A. A. Dia7., J. R. Friley, M. S. Good, Spanner, J. C., S. R. Doctor, T. T. Taylor/PNL and J.
M. S. Greenwood, P. G. Heasler, R. L. Hockey, R. J. Muscara/NRC. 1990. Qualification Process for Ultra-Kurtz, F. A. Simonen, J.C. Spanner, T. T. Taylor, and sonic Testing in Nuclear Inservice Inspection Applica-
- T. V. Vo. 1992. Nondestructive Examination (NDE) tions. NUREG/CR-4882, PNL-6179. Pacific Northwest Reliability for Inservice Inspection of Light Water Reac- Laboratory, Richland, Washington.
tors. NUREG/CR-4469, PNL-5711, Vol. 14. 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, Green, E. R., S. R. Doctor, R. L. Hockey, and A. A. T. T. Taylor, and T. V. Vo. 1990. Nondestructive Ex-Diaz. 1992. Development of Equipment Parameter amination (NDE) Reliability for Inservice Inspection of
- Tolerances for the Ultrasonic Inspection of Steel Compo- Light Water Reactors. NUREG/CR-4469, PNL-5711, nents: Application to Components up to 3 Inches Thick. Vol. 10. Pacific Northwest Laboratory, Richland, NUREG/CR-5817, Vol 1. Pacific Northwest Laborato- Washington.
ry, Richland, Washington.
Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
Green, E. R., S. R. Doctor, J,l. L. Hockey, and A. A. Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, Diaz. The Interaction Matrix Study: Models and Equip- and-T. T. Taylor. 1989. Nondestructive Examination ment Sensitivity Studies for the Ultrasonic Inspection of . (NDE) Reliability for Inservice Inspection of Light Water Thin Wall Steel. NUREG/CR-5817. Pacific Northwest Reactors. NUREG/CR-4469, PNL-5711, Vol. 9. Pacif-Laboratory, Richland, Washington. ic Northwest Laboratory, Richland, Washington.
Doctor, S. R., M. S. Good, P. G. Heasler, R. L. Hock- Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
ey, F. A. Simonen, J. C. Spanner, T. T. Taylor, and Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, T. V. Vo. 1992. Nondestructive Examination (NDE) and T. T; Taylor. 1989. Nondestructive Examination Reliability for Inservice Inspection of Light Water Reac- (NDE) Reliability_ for Inservice Inspection of Light Water tors. NUREG/CR-4469, PNL-5711, Vol. 13. Pacific Reactors'. NUREG/CR-4469, PNL-5711, Vol. 8. Pacif-Northwest Laboratory, Richland, Washington. ic Northwest Laboratory, Richland, Washington.
Doctor, S. R., M. S. Good, P. G. Heasler, R. L. Hock- Doctor, S. R., J. D. Deffenbaugh, M. S. Good, E. R.
ey, F. A. Simonen, J. C. Spanner, T. T, Taylor, and Green, P. G. Heasler, F. A. Simonen, J.C. Spanner, T. V. Vo. 1992. Nondestructive Examination (NDE) and T. T. Taylor. 1988. Nondestructive Examination Reliability for Inservice Inspection of Light Water Reac- (NDE) Reliability for Inservice Inspection of Light Water tors. NUREG/CR-4469, PNL-5711, Vol. 12. Pacific Reactors. NUREG/CR-4469, PNL-5711, Vol. 7. Pacif-Northwest Laboratory, Richland, Washington. ic Northwest Laboratory, Richland, Washington.
Doctor, S. R., M. S. Good, E. R. Green, P. G. Heasler, Doctor, S. R., J. D. De_ffenbaugh, M. S. Good, E. R.
F. A. Simonen, J. C. Spanner, T. T. Taylor, and T. V. Green, P. G. Heasler, G. A. Mart, F. A. Simonen, J.C.
Vo. 1991. Nondestructive Examination (NDE) Reli- Spanner, T. T. Taylor, and L. G. Van Fleet. 1987.
ability for Inservice Inspection of Light Water Reactors. Nondestructive Examination (NDE) Reliability for Inser-NUREG/CR-4469, PNL-5711, Vol. 11. Pacific North- vice Inspection of Light Water Reactors. NUREG/CR-west Laboratory, Richland, Washington. 4469, PNL-5711, Vol. 6. Pacific Northwest Laboratory, Richland, Washington.
xix NUREG/CR-6181
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.
Good, M. S. and L. G. Van Fleet. 1986. Status of Activities for Inspecting Weld Overlaid Pipe Joints . .
Spanner, T. T. Taylor, and L. G. Van Fleet. 1987. NUREG/CR-4484, PNL-5729. Pacific Northwest Labo-Nondestructive Examination (NDE) Reliability for Inser- ratory, Richland, Washington.
vice Inspection of Light Water Reactors. NUREG/CR-4469, PNL-5711, Vol. 5. Pacific Northwest Laboratory, Heasler, P. G., D. J. Bates, T. T. Taylor, and S. R.
Richland, Washington. Doctor. 1986. Perfonnance Demonstration Tests for Detection of Intergranular Stress Co"osion Cracking.
Doctor, S. R., D. J. Bates, J. D. Deffenbaugh, M. S. NUREG/CR-4464, PNL-5705, Pacific Northwest Labo-Good, P. G. Heasler, G. A. Mart, F. A. Simonen, J. C. ratory, Richland, Washington.
Spanner, A. S. Tabatabai, T. T. Taylor, and L. G. Van Fleet. 1987. Nondestructive Examination (NDE) Reli- Simonen, F. A. 1984. The Impact of Nondestructive ability for Inservice Inspection of Light Water Reactors. Examination Unreliability* on Pressure Vessel Fracture NUREG/CR-4469, PNL-5711, Vol. 4. Pacific North- Predictions. NUREG/CR-3743, PNL-5062. Pacific west Laboratory, Richland, Washington. Northwest Laboratory, Richland, Washington.
Collins, H. D. and R. P. Gribble.. 1986. Siamese Imag- *Simonen, F. A. and H. H. Woo. 1984. Analyses of the ing Technique for Quasi-Veltical Type (QVT) Defects in Impact of Inservice Inspection Using Piping Reliability Nuclear Reactor l'iping. NUREG/CR-4472, PNL-5717. Model. NUREG/CR-3869, PNL-5149. Pacific North-Pacific Northwest Laboratory, Richland, Washington. west Laboratory, Richland, Washington.
Doctor, S. R., D. J. Bates, R. L. Bickford, L. A. Taylor, T. T. 1984. An Evaluation of Manual Ultra-
- Charlot; J. D. Deffenbaugh, M. S. Good, P. G. Heasler, sonic Inspection of Cast Stainless Steel Piping.
G. A. Mart, F. A. Simonen, J. 'C. Spanner, A. S. NUREG/CR-3753, PNL-5070. Pacific Northwest Labo-Tabatabai, T. T. Taylor, and L. G. Van Fleet. 1986. ratory, Richland, Washington.
Nondestructive Examination (NDE) Reliability for Inser-vice Inspection of Light Water Reactors. NUREG/CR- . Bush, S. H. 1983. Reliability of Nondestructive Exami-4469, PNL-5711, Vol. 3. Pacific Northwest Laboratory, nation, Volumes I, II, and III. NUREG/CR-3110-1, -2, Richland, Washington; and ~3; PNL-4584. Pacific Northwest Laboratory, Rich-
.land, 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, F. A. and C. W. Goodrich. 1983. Parametric Simonen, J.C. Spanner, A. S. Tabatabai, and T. T. Calculations of Fatigue Crack Growth in Piping.
Taylor. 1986. Evaluation and Improvement of NDE NUREG/CR-3059, PNL-4537. Pacific Northwest Labo-Reliability for Inservice Inspection of Light Water Re- ratory, Richland, Washington.
actors. NUREG/CR-4469, PNL-5711, Vol. 2. Pacific Northwesi Laboratory, Richland, Washington. Simonen, F. A., M. E. Mayfield, T. P. Forte, and D.
Jones. 1983. Crack Growth Evaluation for Small Doctor, S. R., D. J. Bates, L.A. Charlot, H. D. Collins, Cracks in Reactor-Coolant Piping. NUREG/CR-3176, M. S. Good, H. R. Hartzog, P. G. Heasler, G. A. Mart, PNL-4642. Pacific Northwest Laboratory, Richland, F. A. Simonen, J.C. Spanner, and T. T. Taylor. 1986. Washington.
Integration of Nondestructive Examination (NDE) Reli-ability and Fracture Mechanics, Semi-Annual Report,
- Taylor, T. T., S. L. Crawford, S. R. Doctor, and G. J.
April 1984 -September 1984. NUREG/CR-4469, PNL- Posakony. 1983. Detection of Small-Sized Near-Surface 5711, Vol. 1. Pacific Northwest Laboratory, Richland, Under-Clad Cracks for Reactor Pressure Vessels.
Washington. NUREG/CR-2878, PNL-4373. Pacific Northwest Labo-ratory, Richland, Washington.
NUREG/CR-6181 xx:
Previous Reports Busse, L. J., F. L. Becker, R. E. Bowey, S. R. Doctor, Becker, F. L., S. R. Doctor, P. G. Heasler, C. J. Morris, R. P. Gribble, and G. J. Posakony. 1982. Characteriza- S. G. Pitman, G. P. Selby, and F. A. Simonen. 1981; tion Methods for Ultrasonic Test Systems. Integration of NDE Reliability and Fracture Mechanics, NUREG/CR-2264, PNL-4215. Pacific Northwest Labo- Phase I Report. NUREG/CR-1696-1, PNL-3469. Pacif-ratory, Richland, Washington. ic Northwest Laboratory, Richland, Washington.
Morris, C. J. and F. L. Becker. 1982. State-of-Practice Taylor, T. T. and G. P. Selby. 1981. Evaluation of Review of Ultrasonic In-service Inspection of Class I ASME Section XI Reference Level Sensitivity for Initia-System Piping in Commercial Nuclear Power Plants. tion of Ultrasonic Inspection Examination.
NUREG/CR-2468, PNL-4026. Pacific Northwest Labo- NUREG/CR-1957, PNL-3692. Pacific Northwest Labo-ratory, Richland, Washington. ratory, Richland, Washington.
xxi NUREG/CR-6181
'I 1.0 Introduction Pacific Northwest Laboratory (PNL) is conducting a This report describes evaluations for the Surry Nuclear multi-year program for the U.S. Nuclear Regulatory Power Station Unit 1 (Surry-1) which was selected for Commission (NRC) entitled "Evaluation and Improve- demonstrating the risk-based methodology. Participa-ment in Nondestructive Evaluation Reliability for Inser- tion of Virginia Electric Power Company (VEPCO) vice Inspection (ISi) of Light Water Reactors". The staff was an essential part of the pilot study. Plant-goals of this program are to determine the reliability of specific information was obtained through system draw-current ISi of pressure boundary systems and compo- ings, visits to the plant site, and discussions with plant nents, and to develop recommendations that. can ensure operational staff. The specific systems selected for high inspection reliability. The long-term objective is to study were the reactor pressure vessel, reactor coolant, develop technical bases for improvements to the inspec-
- low-pressure injection including accumulators, and the tion requirements of nuclear power plant components. auxiliary feedwater systems. The remaining pressure boundary systems at Surry-1 will be addressed in a Because of similarities in objectives, the PNL program future report. This report presents the results for the is coordinated with the American Society of Mechanical most risk~important components within the four select-Engineers (ASME) Research Task Force on Risk- ed systems at Surry-1 and compares the results for ISi Based Inspection Guidelines. The initial task force priorities with the current ISi practices. Differences are document (ASME 1991) has made general recommen- being assessed to determine the extent of potential dations on the application of risk-based methods to ISi, improvements to ISi plans provided by the new meth-and forms the basis of future proposals to ASME for odology.
improved codes and standards. Results of PNL studies are being made available to the ASME group to dem- Section 2.0 of this report discusses the overall method-onstrate and validate the usefulness of the risk-based ology for risk-based ranking of systems and com-methodology. Future documents specifically addressing ponents. Part of this discussion addresses the methods uclear. power plant components will be issued by the used to estimate component rupture probabilities. An SME Task Force.
- approach is proposed for setting target values for these rupture probabilities at suitable levels. The objective of To provide technical bases for improved ISi plans, PNL this approach is to ensure that the contribution of pres-has developed and applied a method (Vo et al. 1989) sure boundary failures to core damage risk remains a that uses results of probabilistic risk assessments small fraction of total plant risk.
(PRAs) to estimate the consequences of component failures. The probab~ties of these component failures Section 3.0 provides details of the Surry-1 pilot study.
have been estimated by using an expert judgment elici- Descriptions are provided for the four systems tation process (Vo et al. 1991). Using these estimates addressed, and the assumptions made in the analyses .
of consequences and probabilities, risk calculations have are also included. Results of the component rankings established ISi priorities for systems and components at as well the sensitivity and uncertainty analyses are pre-nuclear power plants. Once high-priority components sented. Section 4.0 provides a detailed discussion and have been identified, recommended inspection pro- interpretation of the results of Section 3.0. Finally, a grams (method, frequency, and extent) will be devel- summary and conclusions of the study are presented in
- oped in future work. Probabilistic structural mechanics Section 5.0.
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 NUREG/CR-6181
2.0 Overall Methodology The overall methodology has three major steps: 1) The FMEA results were used .to calculate the impor-selection and risk-prioritization of systems and compo- tance index (or relative importance) for each compo-nents for inspection, 2) selection of a total target risk nent within the selected systems. The importance index value associated with all pressure boundary and struc- was based on the expected consequence of failure of tural failures, and 3) determination of target rupture the component, as measured by the probability of core probability for individual components or structures. damage resulting from component failure. In mathe-The following subsections summarize the overall meth- matical terms, the probability of core damage, Pcm*
odology. resulting from a given component failure (i.e., rupture),
is defined as 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 where Pcm = probability of core damage resulting and components for inspection. Previous work at PNL from the component failure has addressed priorities at a system level as a prelude = failure (rupture) probability of the to component level prioritization. In summary, for a component of interest given system, Iw is defined as the product of the Birn- conditional probability of core dam-baum Importance (I8 ) times the failure probability for age given the failure of system i that system. conditional probability of system i failing given the component failure
= probability that the operator fails to (2.1) recover given a system failure.
I 8 = the change in risk that is associated with Equation (2.2) is totaled for all system failures that a system failure either result directly from the given component failure Pf = system failure probability due to struc- or result indirectly from component damage to other tural integrity failures. compone~ts or the systems in the zone of interest (e.g.,
pipe whip or jet impingement effects, damaging vital Core-damage frequency (Level-I PRA) was used in this electrical buses, etc.). The Standard Review Plan devel-study as the bottom-line risk measure to prioritize the oped by the NRC (1981) and information obtained plant systems. When risk is measured by core damage from plant system walkdowns are used to assess the frequency, the I 8 of a system is equivalent to the condi- indirect effects.
tional probability of core damage given a system failure.
The parameter I 8 is a measure of the consequence of As shown by Equation (2.2), estimates of component structural failure, where Iw also addresses the proba- failure probabilities are required in order to perform bilities of structural failures. Specifically, Iw is an ap- component prioritization. These estimates are summa-proximation of the core damage risk due to system rized in the following subsection.
failures .caused by structural failures. Components with very high I 8 values are given the highest rank in the 2.2 Estimates of Component Rupture risk-based prioritization for inspection planning. Com- Probabilities ponents having high Iw values are then added into this list of highly ranked components. The results of system For each system selected (Table 2.1), the per-compo-prioritization for Surry-1 from an earlier PNL study nent failure probability was estimated. Because histori-(Vo et al. 1990) are shown in Table 2.1. cal failure data on low-probability events (e.g., pipe
-he rupture) are lacking, an expert judgment elicitation was For those systems selected for further analyses, a de- used to estimate component failure probabilities. This tailed component-level prioritization was performed. section summarizes the procedures and the results of FMEA teclmique was selected fo, this analysis.
2:1 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- Each expert then completed questionnaire forms that cussions are given in Vo et al. (1991). addressed location-specific rupture probabilities for the systems of interest. These responses included best The expert judgment elicitation used a systematic pro- estimates of probabilities and uncertainties, and the cedure, which closely followed the approaches reported rationale for these estimates. Following the meetin-,
in the NRC Severe Accident Risks Document (NRC the information provided by the expert panel was re 1989; Wheeler et al. 1989; Meyer et al. 1989). The composed and aggregated. PNL prepared a prelim' specific objective of the PNL elicitation was to develop report of the elicitation, which was then submitted to numerical estimates for probabilities of catastrophic or each panel member for review. This report included disruptive failures in the selected components at Surry- the initial recomposition, additional plant-specific data,
- 1. Figure 2.1 shows information that was used to ob- and other relevant information. The experts were tain the* desired estimates from the experts. requested to review and revise their estimates of rup-ture probabilities. The revised information was again Prior to the expert elicitation workshop, PNL sent recomposed and aggregated to provide single composite reference materials to the experts, including data sourc- judgments for each issue.
es, reports, probabilistic models, and recent PRA re-sults. Panel members were asked to study these mate- Figure 2.2 shows a sample of estimated failure probabil-rials and to make initial estimates of failure probabil- ities obtained from the expert judgment approach.
ities. Similar types of plots were produced for components in other selected systems at Surry-1. For readability, the At the meeting, a formal presentation was provided for probabilities are presented with a log10 scale, with the each system addressed. Presentations covered technical probabilities expressed as failures per component per descriptions, historical component failure mechanisms, year. The ranges of best estimates from the experts elicitation statements, suggested approaches, question- were summarized in a series of plots (boxes and whis-naire forms, and any materials that supported the issue kers) as shown in Figure 2.2. An individual plot dis-descriptions. The presentations were followed by dis- plays five features of the distribution of estimated prob-cussions. The experts provided their knowledge regard- abilities. The "whiskers" display the extreme upper and ing plant design and operation, failure history, material lower bound values of the distribution, while the box degradation mechanisms, and methods for recomposi-. itself locates the 25% and 75% quartiles of the distribu-tion and aggregation of the data. tion. Finally, the circle within the box is the median of the distribution.
NUREG/CR-6181 2.2
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 Discussion
....... (additional plant-specific information, etc.)
~
Estimated Rupture Probabilities
- Figure 2.1. Information Provided to Expert Panel 2.3 Target Risk and Rupture Proba- that the risk of core damage resulting from pressure boundary component structural failures is maintained to bility be less than a small fraction of the total core damage risk estimated by the PRA. The risk due to pressure The purpose of inspections is to keep risk levels within boundary structural failures is referred to as the "target acceptable values (e.g., by detecting and repairing de- risk" and 5% of the total PRA-estimated risk resulting graded components before they lead to rupture). It is a from' internal events has been recommended as an difficult task to select acceptable (or target) values for appropriate numerical value.
the risk associated with pressure boundary component and structural failures. However, once the target risk It is further recommended that this overall target risk values are determined, the corresponding target rup- be apportioned among the risk-important components ture probabilities for individual components can be by considering the risk associated with rupture of each quantified using the methods described in the preceding component. Using the conditional probability of core subsections. This subsection proposes an approach for damage given component failure, the target failure defining the target risk and the target rupture probabili- probability can be calculated for each component from ties. its apportioned share of the overall target risk value.
From this, inspection strategies can be determined to A philosophical approach for selecting target values of maintain component rupture probabilities below target risk and rupture probability for individual components values, and optimum strategies can then be selected.
has been recommended by ASME Research Task Force An example of the target risk and component rupture on Risk-Based Inspection (ASME 1991). This probability calculations is provided in the next section
- approach assumes that the inspection should ensure to clarify the discussion.
2.3 NUREG/CR-6181
2.0 Overall Methodology Legend:
Weld 1 ~ Weld 1 - Circumferential weld, upper Weld2 r-CE}-; shell to intermediate shell
~
Weld 2 - Circumferential weld, flange Weld3 to nozzle Weld4 1---....r--....<p--,~ Weld 3 - Circumferential weld, lower shell (beltline region)
Weld5 <j> ~ Welds 4 - Circumferential welds, Weld6 ~ thru 5 bottom head
~
Welds 6 - Intermediate and lower shell Weld7 thru 9 longitudinal welds (beltline Weld8 <p regions)
Weld9 --1
Welds *1 O -
thru 15 Nozzle to vessel welds Weld 10 ~ CRDMs - Control rod drive mechanisms Weld 11 ~ Upper/Lower Bound Weld 12 ~ ....__ _.l 25%175% Quartile Weld 13 ~ <j> Median
~
Weld 14 Weld 15
~
Nozzle Forgings - Inlet cj> i--l ------1 Nozzle Forgings - Outlet <p i - -
Beltline Plate
~
Vessel Shell - Outside Beltline
~
Upper Head I ~~
Lower Head
~ ~
Vessel Flange I ~ I Enclosure Head Flange I ~ ~
Vessel Studs ~ <p ~
~
Instrument Line Penetrations
~
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 tion obtained from the walkdowns was later used to Surry-1 systems. Identification and prioritization ~f assess the indirect effects on the systems. The walk-components for the Surry-1 plant systems are provided downs for each system included the plant engineer and following a brief discussion of plant familiarization, one or two project team analysts. For each component system descriptions, and analysis assumptions. The (e.g., pipe segment), all the necessary information relat-section concludes with sensitivity analyses. ed to that component was obtained. This information was entered into the preliminary FMEA niodels. For 3.1 Plant Familiarization example, for a given pipe segment within a selected system, the component identification, including the pipe Participation of VEPCO was an essential part of the size, W<!,S identified. Ntimbers of welds, elbows, sup-pilot study. Before initiating the pilot study, a visit to ports, connections, penetrations, etc., within the pipe VEPCO headquarters was conducted. The purpose of segment in question were identified and recorded.
this first visit was to get acquainted with VEPCO per- Given a component failure, the potential targets that sonnel and to request needed data. might be impacted by the failed components ( e.g., vital electrical buses, system components nearby, etc.) were Prior to the initial plant visit, the project team analysts also recorded. Additionally, a video camera was used reviewed the fault trees reported in the Surry-1 PRA, to record the conversations with the responsible engi-the system descriptions, and the sections of the final neer and views of significant locations of concern to sys-safety analysis report (FSAR) applicable to the systems tem design and operation as identified during the initial of interest. The preliminary FMEA models were con- visit.
structed and preliminary success criteria and dependen-cy matrices were developed to identify specific areas In addition to the plant system walkdowns, discussions here information was needed to develop an accurate with plant operational and technical staff were also odel. Based on these initial activities, a letter of conducted. The areas of discussion included plant and quest was prepared and sent to the plant to identify system modeling questions, collections of system design the plant-specific information and data that was and operational information, discussions of transient required. The following subsections provide a descrip- sequence progressions, and the operators' responses to tion of the plant visit and the information obtained these events. During the plant visit the team had dis-during the visit. 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, 3.1.1 Initial Plant Visit the shift technical advisor, and members of the mainte-nance and engineering staff.
A one-week plant visit was arranged to meet with plant personnel. During this visit, project team analysts Discussions centered on gaining a clear understanding performed the system walkdowns and obtained relevant of the following items:
plant information. The visiting PNL team included plant system specialists and PRA specialists. Because the plant was in operation during the initial visit, system
.. the normal and emergency configurations and operations of the various systems of walkdowns for some locations were not possible (e.g.,
interest inside the containment building and other high-radia-tion areas). Therefore, the Video Information Manage-ment System (VIMS) developed by VEPCO was also
- system dependencies used. VIMS is a computerized system, that displays photographs of plant systems and components that have
- operational problem areas identified by plant personnel that may impact the analysis 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.
- automatic and manual actions taken in re-sponse to various emergency conditions For each of the systems selected for the study, a system alkdown was conducted where possible. The informa-
- availability of plant specific operational data .
3.1 NUREG/CR-6181
3.0 Analyses The emergency procedures which addressed actions identified by the project analysts as important actions were explained to operations personnel.
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-3.1.2 Information Obtained inghouse-designed, three-loop, pressurized-water reactor (PWR) rated at 788 MWe capacity with a sub-atmo-A complete set of the current Surry piping and instru- spheric containment. The balance of the plant and mentation drawings (P&ID), isometric drawings, com- containment building were designed and constructed by posite drawings, and stress analysis reports were provid- Stone and Webster Engineering Corporation. Surry-1 is ed by the Surry-1 staff. Also, the Surry-1 staff provided operated by VEPCO. Commercial operation started in copies of the Surry Emergency Procedures, Abnormal 1972.
Procedures, Emergency Contingency Action Proce-dures, Functional Restoration Procedures, and several The Surry-1 systems selected for study were the primary sections from the current revisions of the Surry-1 pressure boundary system, the front-line safety systems, FSAR. The plant information was incorporated into and certain important support systems identified in PNL's preliminary FMEA models. For instance, the Table 2.1. These Were the reactor pressure vessel isometric and composite drawings were used to obtain (RPV), reactor coolant (RCS), low-pressure injection additional information regarding component orientation (LPI), and the auxiliary feedwater (AFW) systems. The and number of subcomponents. The Emergency Proce- following paragraphs summarize the descriptions for dures were used to assess the recovery actions by the these systems. Detailed descriptions can be found in operators given a rupture of component. the Surry-1 FSAR.
3.1.3 Subsequent Plant Visits 3.2.1 Reactor Pressure Vessel During the course of the study, two additional plant The RPV is a principal component of the RCS.
visits were conducted. The first visit was to obtain ad- Surry-1 RPV is shown in Figure 3.1. It consists of a ditional plant-specific failure mechanisms for compo- cylindrical shell with a hemispherical bottom head, and nents within the system analyzed. This information was a flanged and gasketed removable upper head. The provided to an expert workshop on estimating compo- vessel contains the core, core support structures, control nent rupture probabilities. The other plant visit was rQds, thermal shield, and other parts directly associated conducted during the plant shutdown for refueling. with the core. Outlet and inlet nozzles are located This visit was to obtain additional information and to between the upper head and the core.
verify the information that was obtained from an initial visit (e.g., areas inside the containment building). PNL The Surry-1 vessel was designed and manufactured by is currently performing the pilot study for the Surry-1 the Babcock and Wilcox Company to the requirements balance of the plant system and additional plant visits of Section Ill of the ASME Boiler and Pressure Vessel are anticipated. Code (BPVC). Design features and materials selection are typical for PWR reactor vessels at U.S. nuclear 3.1.4 Utility Interface plants. The vessel is designed of low-alloy steel with forgings of Type A508, Class 2 and plate materials of An ongoing interface was maintained with the utility Type A533, Grade B, Class 1. All surfaces in contact throughout the duration of the analysis. The project with coolant are clad with, or made from, 300-series team leader was in frequent contact with Surry-1 plant stainless steel or Inconel. In general, all attachments personnel to ask questions and verify information. and pressure-containing parts have full-penetration Surry-1 personnel also reviewed the results of the study welds. Partial welds are used to attach the relatively when they became available. small diameter control rod drives and the instrumenta-tion tubes to the vessel heads.
NUREG /CR-6181 3.2
- 3.0 Analyses Flange Ligaments 1 thru 58 (Ref, 1-1100A) 2 _... ,.....t---.-------,r-------l~ Diameter: 157.0" I.D.
Circumference: 492.98" I Material: Flange-A508 Class
~ 6 7 2 Carbon Steel (45°} (225°}
Upper Shell: 9.125"T - 508 Class 2 Carbon Steel 3--..- Intermediate Shell: 9.0"T -
A533 Carbon Steel I
8~ Lower Shell: 9.0"T - A533 (135°} Carbon Steel 9
(315°}
4--..-
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 three identical heat transfer loops (connecting parallel to the RPV), each of which includes a steam generator, The function of the RCS is to remove heat and transfer reactor coolant pump, and connecting piping and instru-it to the secondary system. It also provides a barrier mentation for flow and temperature measurements.
against the release of reactor coolant or radioactive materials to the containment environment. ' The RCS The pipes through which the heated water flows from for Surry-1 is diagramed in Figure 3.2. It consists of the RPV to the steam generator are called the "hot legs" and the pipes through which the cooled water 3.3 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 coolant loop and is maintained at the saturation tem-are called the "cold legs." The working fluid is boiled perature that corresponds to the system pressure.
on the secondary sides of the steam generator and transported through a conventional turbine-condenser To regulate the reactor coolant chemistry within design system. limits and control the pressure level, a constant letdown flow from one loop upstream of the reactor coolant The RCS also includes a pressurizer that maintains the pump is maintained. This flow is, in turn, controlled by reactor coolant at a constant pressure. The pressurizer the pressurizer level. Constant coolant makeup is add-system consists of power-operated relief valves with ed by charging pumps in the chemical and volume associated block valves, ASME code safety valves, pres- control systems. The inservice integrity of the RCS is surizer sprays, and electrical heaters. There is continu- addressed through periodic inspections performed in ous control of the water and steam inventory within the accordance with the requirements of ASME, Section pressurizer vessel. The pressurizer is connected to a XI.
NUREG /CR-6181 3.4
- 3.0 Analyses 3.2.3 Low-Pressure Injection System The accumulators, which are passive components, serve as another injection mode for the LPI system .. They
. The LPI consists of several independent subsystems provide an initial influx of borated water to reflood the characterized by equipment and flow path redundancy reactor core following a large or medium LOCA. The inside the missile protection boundaries. The two phas- accumulator system, diagrammed in Figure 3.4, consists es of low-pressure system operation including active of three tanks filled with borated water and pressurized low-pressure injectiqn and recirculation mode and the with nitrogen. Each of the accumulators is connected passive accumulator injection are summarized below. to one of the RCS cold legs by a line containing*a normally open MOV and check valve in series. The The Surry-1 low-pressure injection/recirculation system check valves serve as isolation valves during normal (LPI/LPR) provides emergency coolant injection and operation and open to empty the contents of the accu-recirculation following a loss-of-coolant accident mulators when the RCS pressure falls below 650 psig.
(LOCA) when the RCS depressurizes below the low-pressure setpoint (about 300 psig). In addition to the The accumulators depend on the nitrogen system to direct recirculation of coolant during the recirculation maintain the pressure head. The nitrogen is supplied phase once the RCS is depressurized, the LPR dis- by dedicated local nitrogen bottles, and the accumula-charge provides the suction source for the high-pressure tors are fully instrumented to indicate abnormal pres-recirculation system following drainage of the refueling sure conditions. The accumulators are initially filled water storage tank (RWST). with borated water storage from the RWST, and the valves are dosed. Instrumentation verifies that the level
- The LPI/LPR at Surry-1 is diagrammed in Figure .3.3. remains above a minimum value.
The system consists of two 100% capacity pump trains.
In the injection mode, the pump trains share a common
- The associated components, piping, structures, and uction header from the RWST. Each pump draws power supplies of the LPI system (including the accu-suction from the header through normally open motor- mulators) are designed to conform with Class 1 seismic
. operated valves (MOVs), check valves, and locked-open criteria. All motors, instruments, transmitters, and their manual valves. Each pump discharges through a check associated cables located inside the containment are valve and normally open MOV in series to a common designed to function during and under the postulated injection header. The injection header contains a temperature, pressure, and humidity conditions.
locked-open MOV and branches to separate lines, one to* each cold leg. Each of the lines to the cold legs All LPI piping in contact with borated water is austenit-contains two check valves in series to provide isolation ic stainless steel. The piping is designed to meet the from the high-pressure RCS. minimum requirements set forth in B31.l Code for Pressure Piping, B36.10 and B16.19, ASTM Standards, In the recirculation mode, the pump trairis draw suction Supplementary Standards, and Additional Quality Con-from the containment sump through a parallel arrange- trol Measures. The piping is supported to accommo-ment of suction lines to a common header. Flow froni date expansion due to temperature changes and hydrau-the suction header is drawn through a normally closed lic forces during an accident. All components of the MOV and check valve in series. Discharge of the LPI/LPR and accumulators are tested periodically to pump is directed to either the cold legs through the demonstrate system readiness. All pressure piping butt same lines used for injection or to a parallel set of welds containing radioactive fluid, at greater than 600°F headers that feed the charging puinps, depending on and 600 psig, were radiographed. The remaining piping the RCS pressure. butt welds were randomly radiographed. Pressure-containing components are inspected for leaks from In the hot-leg injection mode, system operation is iden- pump seals, valve packing, flanged joints, and safety tical to normal recirculation with the exception that the valves during system testing. Frequency of testing and normally open cold-leg injection valves must be manual- maintenance of the system components are specified in ly closed remotely, and one or more normally closed the ASME,Section XI.
hot-leg recirculation valves must be manually opened.
- 3.5 NUREG/CR-6181
To Charging Pump Inlet Header From HPI NC-FAI 6"-Sl-50-1502 To Hot Leg Loop 3 HPI "--
PS33 Hot Leg Loop 2 CV50 1890B CV228 From HPI XV48 (1-SI-P-18) NOFAI MDPSIIB 1864D Cold Leg Loop 1 PS35 Power Removed CV241 CV79 NOFAI PS44 6"-Sl-152-1502 Cold Leg Loop 2 PS36 CV242 CV82 PS34 1890C NOFAI PS45 D 1864A Cold Leg Loop 3 CV243 CV85 6"-Sl-49-1502 Hot Leg Loop 1 1890A CV229 (1-SI-P-1 A) 1863A PS46 PS32 PS39 MDPSIIA From HPI Sump To Charging Pumps AL - Out of Position Alarm in Control Room R9312053.3 Figure 3.3. Surry-1 Low-Pressure Injection/Recirculation System Simplified Schematic
3.0 Analyses 1-S1-TK-1A From RWST FC Loop 1 CV107 CV ~ -
109
1 Cold Leg 1865A 1-SI-TK-1B Loop2 FC ,.1-----....;r,1------1 Cold Leg 1865B CV128 CV130 1-S1-.TK-1C
.FC 1865C CV145 CV147 R9312053.1 Figure 3.4. Surry-1 Accumulator System Simplified Schematic 3.2.4 Auxiliary Feedwater System discharges to parallel headers; each of these headers can provide AFW flow to any or all of the steam gener-The AFW system provides feedwater to the steam ators. Flow from each header to any one steam gener-generators for heat removal from the primary system ator is through a normally open MOV and locked-open after a reactor trip. The AFW system may also be used valve in series, paralleled with a line from the other following a reactor shutdown, in conjunction with the header. These lines* feed one line containing a check condenser dump valves or atmospheric relief valves, to valve that joins the main feedwater line to a steam cool the RCS to about 300°F and 300 psig, at which generator.
time the residual heat removal system is brought into operation. The AFW system also provides emergency The motor-driven pumps automatically start on receipt water following a secondary-side line rupture. Removal of a safety actuation system signal, loss of main feed-of heat in this manner prevents the reactor coolant water, low steam generator level in any steam genera-pressure from increasing and causing release of reactor tor, or Joss of off-site power. The turbine-driven pumps coolant through the pressurizer relief and/or safety automatically start on indication of a low steam genera-valves. tor level in any steam generator or undervoltage of any of the main RCS pumps. '
The AFW system is diagramed in Figure 3.5. The AFW is a multiple-train system; it consists of electric Most of the AFW equipment is located in the auxiliary motor-driven pumps and steam turbine-driven pumps. building. This building is designed to withstand the Each pump draws suction through an independent line effects of earthquakes, tornadoes, floods, and other from the condensate storage tank. Each .(',FW pump natural phenomenon. Provisions are incorporated in 3.7 NUREG/CR-6181
_ _ _ _ _ _ _..,.._ Main Steam XN87 u, 0 LO
'<t
'<t
(.) X X
300,000 GAL CST XV120 LO PS95 PS96 Header B CV178 Header A CV182 To Unit 2 AFW System w
bo PS94
.-- - MOVFW260A
..........~ ..... MOVFW260B ADVMS102B ADVMS102A XN'ZTO XV271 Turbine Drive for PS84 CV133 CV131 PumpTDPFW2 PS83 CV138 MOVFW160B
_..,_...,....,...,...,.....,._. ...,.,... . From Fire Main CV309 From Unit 2 AFW Pumps
,J--.....~~-
- ..............,""""-""'"'16-""' From Emergency Makeup System CV273 MOVFW160A R9312053.2 Figure 3.5. Surry-1 Auxiliary Feedwater System Simplified Schematic
3.0 Analyses the AFW design to allow periodic operation to dem- the system* pressure boundary, are not con-onstrate performance and structural leak-tight integrity. sidered. However, failures of these compo-Leak detection is provided by visual examination and nents as pressure boundaries are addressed.
sensors in the floor drain system. The capability to . Steam generator tube failures have been isolate components or piping is provided, if required, so considered in other studies and are not in-that the AFW system's safety function will not be com- cluded in this study.
promised. Provisions are made to allow for ISi of components at. the appropriate times specified in the
- The Standard Review Plan 3.6.2, developed by the ASME,Section XI. -NRC (1981), was used in determining the indirect effects (e.g., pipe whip, jet forces, etc.) of compo-3.3 Analyses Assumptions nent failures, as such failures relate to other com-ponents in the zone of interest ( e.g., vital electri-General assumptions used for the analyses are the cal buses). Additionally, when a larger diameter following: pipe impacts a smaller diameter pipe of the same pipe schedule, a smaller diameter pipe is assumed
- Core damage frequency was used as the to fail.
bottom-line risk measure to prioritize plant system components.
- Potential flooding due to pipe ruptures that could damage safety-related systems and
- For the four selected systems, the discrete equipment are not included in these analy-components* (piping segments, welds, etc.) ses. Floodings will be addressed at the later are identified for purposes of the risk-based date.
evaluation. For the RPV, the major compo-nents of interest were the vessel shell, heads, 3.4 Component Prioritization flanges, closure studs, penetrations, nozzles and safe ends, and attachment welds. For The quantitative FMEA technique,. as described in other systems, the coinponents of interest Section 2.0, was used to prioritize components on the were pipe segments. These included the basis of core damage risk. In summary, the following straight lengths of pipe, pipe elbows, cou- sources of information were used to prioritize compo-plings, fittings, flanged joints, and welds. nents for inspection: 1) the component failure proba-Additionally, tanks and heat exchangers, bilities estimated from expert judgment elicitation (Vo including the pressurizer, are also included et al. 1990), 2) the results froni Surry-1 system prioriti-as components in the analyses. zation (Vo et al. 1989), and 3) system fault trees report-ed in the Surry-1 PRA (Bertucio and Julius 1990). The
- The system Birnbaum Importance results *Integrated Reliability and Risk Analysis System were used to provide the conditional proba- (IRRAS) computer program developed by Idaho Na-bilities of core damage given the system tional Engineering Laboratory (Russel et al. 1987) was failures. used to reanalyze the developed fault trees ( e.g., calcu-late the conditional probability of system failure given a
- Identical components in identical trains component failure).
within the same system were assumed to have the same failure consequences. The FMEAs were initially formulated using plant sys-tem drawings and other relevant plant-specific infor-
- In these analyses, failures in piping of less mation. As stated in the assumptions, Standard Review than 1-in. in diameter generally are not Plan information developed by the NRC was used in considered, primarily because of the enorm- determining the potential effects of system component ous amount of instrumentation piping of this . failures on other components in the zone of interest.
size. Active functions of components such To ensure that plant models were as realistic as possi-as pumps and valves, which make up part of ble and reflected plant operational practices, visits to 3.9 NUREG/CR-6181
3.0 Analyses the Surry-1 plant were conducted for plant system walk- probability of core damage, given the vessel failure, downs, and discussions were held with plant operational Pcm Is, was assumed to be 1.0.
and technical staff. For locations where the walkdowns were not possible, (e.g., high-radiation areas) the VIMS Depending on failure location and/or accident scenario, developed by VEPCO was used to identify the poten- the recovery action, Ri, was assigned an estimated tially impacted systems and equipment (given a failure probability based on discussions with the plant technical of a component in the zone of interest). staff and on information obtained from the PRA. For the reactor pressure, no recovery action was possible.
The FMEA worksheets were devised so that the neces- It is important to note that in this study the probability sary information could be systematically tabulated. In of recovery by the operator staff was incorporated in the following paragraphs, the example of the RPV the system prioritization scheme. To prevent double (using Equation 2.2) is discussed. Copies of FMEA counting, the recovery actions were assessed qualitative-worksheets are provided in Appendix A of this report. ly. For each postulated failure of a component within the selected systems, the core damage probability was The first step of the analysis was to identify the compo- calculated. A computer program was developed for the nent locations and/or the number of subcomponents calculations.
within a specified pipe segment or region. For exam-ple, the beltline region of the Surry-1 RPV consists of The product of the component failure probability and five welds (four longitudinal welds and one circumfer- the corresponding core damage probability given a ential weld). The per-weld failure probability, Pr, was failure of the component, was calculated. This value estimated as 3.2E-07 (see Figure 2.2). The failure describes the expected risk-based implication of the probability. of the beltline region (five welds) was esti- component under consideration. In the vessel example, mated as 5* (3.2E-07 /weld) = 1.6E-06. failures of welds at the vessel beltline region were as-sumed to result in loss of the vessel, and the core dam-As discussed in a previous section, the consequences of age probability per plant year was estimated to be component failures were to be placed into two catego- (1.6E-06)
- 1 = i.6E-06.
ries, those that resulted in direct effects on the system in question and those that resulted in effects on other On the FMEA worksheets the relative importance of systems or components in the zone of interest (e.g., each component was calculated as illustrated above component failures due to pipe whip or jet impinge- (e.g., 1.6E-06
- 1). An importance index was used to ment effects). In either case, the total contribution to rank each component in a given system by normalizing core damage, given a failure of the component under its core damage probabilities to that of a component consideration, were assessed ( e.g., the product of the with the highest core damage probability. The highest conditional probability of core dam*age given system value of the index identifies the component that is the failures and the probability of system failures given most important for the system being analyzed. A final component failures, Pcm Is
- P6 IP). combined ranking of components for all four systems together was developed based on the numerical values Information from prior system level prioritizations, for core damage frequency.
system walkdowns, discussions with VEPCO staff, the Standard Review Plan, and the fault trees reported in 3.5 Results of Analyses 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- Within the four systems analyzed, there are approxi-nent failure. The IRRAS computer program was used mately 250 major components (or pipe segments). By to calculate probabilities of system failure given a com- assuming that identical components in identical trains ponent failure. Generally, the Birnbaum Importance within the same system have the same failure probabili- .
Measure for the system was used to provide the condi- ties and consequences, these components are reduced tional probability of core damage given a system failure. to approximately 125 components. For ranking pur-For the RPV, the primary effect of a weld failure was pose, components within the same train can be further the loss of the vessel (P s IP = 1.0). In this case, the grouped, based on major discontinuities (e.g., between NUREG/CR-6181 3.10
3.0 Analyses pumps and major valves). This resulted in 37 major 3.6 Sensitivity and Uncertainty Analy-component groups within the systems analyzed.
ses Table 3.1 shows the results of the risk-based ranking of major components within the four selected systems at There are various sources of uncertainty in the numeri-Surry-1, based on the contributions of component fail- cal results of this study. This section describes specific ures to core damage frequency. Included in the table sources of uncertainty .and provides the results of uncer-are the upper- and lower-bound values estimated for tainty/sensitivity analyses.
each component to indicate the effects of uncertainties in the estimates of component rupture probabilities. 3.6.1 Treatment of Uncertainties The rankings ( as shown in the table) are based on the median values estimated from the Surry-1 PRA and Two basic types of uncertainties addressed in this study PNL evaluations of other factors such as rupture proba- were parameter value uncertainty and modelqig uncer-bilities, as discussed in the preceding section. Fig- tainty. Parameter value uncertainties were evaluated ure 3.5 presents this information graphically for the for component rupture probabilities, the conditional components in the four systems. probability of core damage* given component failures, and human recovery action probabilities. Modeling As shown in Table 3.1, the contributions of individual uncertainty was evaluated for the treatment of the indi-component failures to core damage frequency (based on rect effects of the component failures.
the median values) range widely from about 1.6E-14 to 1.58E-06 per plant year. The cumulative risk contribu- The uncertainties of the component rupt~re probabili-tion from all components as shown in Figure 3.6 is ties have been addressed in Vo et al. 1990. For exam-about 2.lE-06 per plant year. Figure 3.7 shows this
- ple, the population quartile was chosen to describe cumulative risk contribution. It is interesting to note uncertainty in the estimates of component rupture that the risk contribution is domina:ted by approximately probabilities (see Figure 2.2). Limited uncertainty the first 18 highest-ranked components. The system analyses regarding the core damage conditional proba-level rankings obtained by summing component contri- bilities have been addressed. The uncertainties in com-butions are the following: 1) RPV, 2) LPI, 3) AFW, ponent unavailabilities, initiating event frequencies, and and 4) RCS. These system level rankings agree with cut set element unavailabilities and their associated those obtained in an earlier PNL study (Vo et al. 1989). modeling were not addressed in this study. Consider-ation of functional dependencies and common-cause For the purpose of comparison, the components are effects on systems were based on the results evaluated also ranked on the basis of the .calculated values of
- by the selected PRAs. The mean parameter values conditional probability of core damage given a compo- estimated by the PRAs were used to calculate the core nent failure. Table 3.2 shows this ranking based on the damage conditional probabilities. The uncertainties of median values estimated. Figure 3.8 presents this same recovery action errors were addressed in the Surry-1 information graphically. The conditional contributions PRA (Bertucio and Julius 1990). In these evaluations of component failures to core damage range from 1.0 to the probabilities were assessed using values of each about 1.0E-07. As expected, the highest contributions parameter such that the nth percentile (or quartile) of are from RPV components, since rupture of the RPV the uncertainty distribution representing the range over beltline region leads directly to core damage. Condi- which the true values lie. In this, Monte Carlo or other tional contributions to core damage from the LPI, the sampling techniques were used to assess the propaga-AFW, and the RCS components are lower due to the tion of parameter value uncertainty through the final ability of redundant safety systems to mitigate accidents results.
and, hence, prevent core damage. Table 3.3 shows the risk importance parameters for 37 major components There are many variables involved in calculating the identified in Table 3.1. The component rupture fre- indirect effects given a component failure (e.g;, location quencies (as shown in Table 3.3) were the average of of pipe break, orientation of the equipment, direction of component group. whipping pipe, number of hangers and/ or supports, impact location, angle of impacts, etc.). Guidance 3.11 NUREG /CR-6181
3.0 Analyses provided in the Standard Review Plan 3.6.2 and infor- the indirect effects to the overall core damage frequen-mation obtained through discussions with VEPCO staff cy are negligible (less than 2%). Two pipe segments during system walkdowns were used to assess the indi- identified to have potential failure effects on the other rect effects. The assessments of the indirect effects systems nearby were 1) the pipe segment between the using Standard Review Plan 3.6.2 are likely to be con- accumulator discharge line and RCS isolation valve and servative. The uncertainty was evaluated by excluding 2) the pipe segment between LPI pump discharge line the potential indirect effects of component failures in and the containment isolation valve. Rupture of the the model (e.g., pipe whip or jet impingement effects) accumulator discharge line could result in a failure of and recalculating the overall core damage frequency. the entire residual heat removal discharge line due to its potential pipe whip and/or jet impingement effects.
3.6.2 Results of Uncertainty/Sensitivity Similarly, rupture of the LPI pump discharge line could Analyses
- result in disabling of the charging pump inlet header.
Sensitivity analyses were performed on issues that could Although analyses regarding potential flooding within potentially have significant impact on component rank- the plant due to pipe ruptures were not part of this ing. The sensitivity analyses addressed the changes in study, the safeguard room at Surry-1 needs to be men-component rankings by using upper- and lower-estimat- tioned. The safeguard room houses the AFW pumps ed values of component rupture probabilities as report- and serves as a pass-through area for the main steam ed in calculations in Vo et al. (1990). As shown in *lines and the main feedwater lines. The room also Table 3.1, although variation exists in the numerical contains three main steam isolation valves, three main results, most components have relatively the sanie rank- steam valves, fifteen steam generator atmospheric dump ing, as compared to the ranking based on the median valves, three steam generator relief valves, one small values. The largest variations in component ranking decay heat relief valve, and three main feedwater check
- were the LPI supply lines and sources, pipe segments valves. The concern is that a rupture in a pipe seg-extending from isolation valves to the steam generator, ment, a valve body, or a steam-water line could flood and pump suction and discharge lines of the AFW *
- the room with steam and/or water, thereby causing all system. Pipe segments between the RPV and the RCS. AFW pumps and other safety-related equipment to fail.
loop stop valves have moderate variations in ranking. This is an important issue and will be addressed at the
- later date. Component risk prioritization of the entire Sensitivity analyses were also performed by letting the
- reactor system will be completed and the main steam component rupture probabilities approach 1.0. This and main feedwater lines will also be evaluated at a causes Pcm values to be the same as Pcm IPs** The new later date.
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 3.12
3.0 Analyses Table 3.1. Component Rankings Based on Core Damage Frequency for Four Selected Systems at Surry-l(a)
Core Damage Frequency System - Component(b) Rank(c)
Upper Median Lower RPV- Beltline Welds 4.00E-04 1.58E-06 5.40E-07 1 RPV- Beltline Plate 1.00E-07 1.00E-07 8.00E-08 2 RPV~ Lower /Bottom Shell 2.SOE-07 7.32E-08 2.00E-08 3 AFW- CST, Supply Line 1.60E-07 6.86E-08 1.60E-08 4 RPV- Circumferential Flange to Nozzle 1.30E-07 6.16E-08 2.00E-08 5 Course, Upper Shell, Outside Beltline Welds LPI-A - Pipe Segment Between Accumulator i.60E-07 4.67E-08 3.20E-09 6 Discharge Header and RCS Isolation Valves LPI - Pipe Segment Between Containment 1.60E-07 4.16E-08 3.2E-08 7 Isolation Valve (inside) and Cold Leg Injection LPI - Pipe Segment Between Containment 8.70E-08 3.80E-08 1.40E-08 8 Isolation Valve (inside) and Cold Leg Injection LPI - LPI Sources (RWST, Sump), Supply 1.86E-07 3.64E-08 3.20E-10 9 Line LPI- Pipe Segment Between Pump Dis- 8.64E-08 2.76E-08 8.80E-09 10 charge and Containment Isolation Valve LPI - Pipe Segment Between Containment 5;44E-08 1.46E-08 8.00E-09 . 11 Isolation Valves RPV- CRDMs 2.74E-08 5.00E-09 2.23E-09 12 RPV- Instrument Lines 4.18E-08 5.00E-09 1.58:E-09 12 AFW- Pipe Segment Between Containment 8.00E-09 3.33E-09 1.52E-10 14 Isolation and SG Isolation Valves AFW- Main Steam to AFW Pump Turbine 8.36E-09 2.lOE-09 2.79E-10 15 Drive RCS - Pipe Segment Between Loop Stop 2.70E-09 1.60E-09 5.34E-10 6 Valve and RPV (Cold Leg) 3.13 NUREG/CR-6181
3.0 Analyses Table 3.1 (cont'd)
Core Damage Frequency System - Component(b) 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-RPV-AFW-AFW-Upper, Closure Head, Flange Nozzle Forging Inlet/Outlet AFW TDP Suction. Line Pipe Segment from Emergency Make-5.71E-11 2.SOE-11 4.37E-11 2.46E-11 3.58E-11 2.SOE-11 1.SlE-11 5.71E-12 5.71E-12 5.00E-12 5.41E-12 2.46E-12 27 28 29 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
- 3.0 Analyses Table 3.2. Component Rankings Based on Conditional Core Damage Frequency Given a Component Rupture<a) for Selected Systems at Surry-1 Core Damage System-Component(b) Frequency Rank RPV- Beltline Welds 1.0 1 RPV- Beltline Plate 1.0 1 RPV- Lower /Bottom Shell 1.0 1 AFW-. CST, Supply Line l.70E-02 9 RPV- Circumferential Flange to Nozzle Course, Upper Shell, 1.0 1 Outside Beltline Welds LPI-A - Pipe Segment Between Accumulator Discharge Header and l.SOE-02 8 RCS Isolation Valves
- LPI - Pipe Segment Between Containment Isolation Valve (inside) l.60E-02 10 and Cold-Leg Injection PI - Pipe Segment Eetween Containment Isolation Valve (inside) 3.00E-02 6 and Hot-Leg Injection LPI - LPI Sources (RWST, Sump), Supply Line 3.64E-02 5 LPI - Pipe Segment Between Pump Discharge and Containment 2.00E-02 7 Iso. Valve LPI - Pipe Segment Between Containment Isolation Valves l.20E-02 11 RPV- CRDMs 5.00E-04 19 RPV- Instrument Lines 5.00E-04 19 AFW- Pipe Segment Between Containment Isolation and SG Isola- 4.00E-04 27 tion Valves AFW- Main Steam to AFW Pump Turbine Drive l.60E-04 31 RCS - Pipe Segment Between Loop Stop Valve and RPV (Cold l.lOE-02 12 Leg)
LPI - LPI Pump Suction Line l.40E-03 17 RCS- Pressi.µ-izer Spray Line l.OOE-04 18 RCS - Pipe Segment Between RPV and Loop Stop Valve (Hot Leg) 2.90E-03 14 AFW- AFW TD Pump Discharge Line 2.30E-04 30 AFW- Pipe Segment from Unit 2 AFW Pumps l.40E-04 32 3.15 NUREG /CR-6181
3.0 Analyses Table 3.2 (cont'd)
Core Damage System-ComponentCb) 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
3.0 Analyses Table 3.3. Risk Importance Parameters for Components at Selected Systems at Surry-1(a)
Conditional Core Damage Core Frequency Rupture Damage System-Component(b) Rank Given Rupture Frequency Frequency RPV - Beltline Region Welds 1 1.0 1.58E-06 l.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 1.7E-02 4.03E-06 6.86E-08 RPV - Circumferential Flange to Nozzle 5 1.0 6.16E-08 6.16E-08 Course Upper Shell, Outside Beltline .-
Welds LPI-A - Pipe Segment Between Accumu- 6 1.8E-02 *2.s9E-06 4.67E-08 lator Discharge Header and RCS Isola- -----:---
tion Valves PI - Pipe Segment Between Contain- 7 3.2E-02 1.30E-06 4.16E-08 ment Isolation Valve (inside) and Cold Leg Injection LPI - Pipe Segment Between Contain- 8 3.20E-02 1.19E-05 3.80E-08 ment Isolation Valve (inside) and Hot Leg Injection LPI - LPI Sources (RWST, Sump), Sup- 9 3.64E-02 1.00E-06 3.64E-08 ply Line LPI - Pipe Segment Between Pump Dis- 10 3.2E-02 8.63E-07 2.76E-08 charge and Containment Isolation Valve LPI - Pipe Segment Between Contain- 11 1.6E-92 9.13E-07 l.46E-08 ment Isolation 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 Contain- 14 8.49E-05 3.92E-05 3.33E-09 ment Isolation and SG Isolation Valves AFW - Main Steam to AFW Pump Tur- 15 1.64E-04 1.28E-05 2.lOE-09 bine Drive RCS~ Pipe Segment Between Loop Stop 16 1.13E-02 1.42E-07 1.60E-09 Valve and RPV (Cold Leg) 3.17 NUREG/CR-6181
3.0 Analyses Table 3.3 (cont'd)
Conditional Core Damage Core Frequency Rupture Damage System-Component(b) Rank Given Rupture Frequency Frequency LPI - LPI Pump Suction Line 17 1.36E-03 l.10E 06 0
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
10*5 10*6 10*1 25 and 75 { AMedian Quartiles 1 Value 10-s u
C G) 0 ~
er 10*9 C G) 0 ..
'S u.G)
.Q C, .10-10 w
- ... *- ca
\0 .t:: E C ca OQ o! 10-11 0
0 10-12 1 2
RPV*Bsltline Region Welds RPV*Beltllne Plate 11 LPl*Plpe Segment Between Containment Isolation Valves 22 23 AFW-lsolatlon Va!ve lo SG RPV-Nazzfe to Vessel Wekts 3 RPV*Lower/Bottom SheU 12 RPV,CRDMS 24 RPV-Vessel Studs 2J}J 4 AFW*CST, Supp/'y Line 13 RPV*lnstrument Unes 25 AFW-AFW MOP SUctlon Uno 5 RPV-Clr. Flange to Nozzle Course, 14 AFW*Plpe Segment Between Containment 26 AFW-AFW MOP Discharge Line Upper Shen. Outside Beltllne Welds Isolation & SO Isolation Valves 27 RPV*Upper, Closure Head Flange LPl*A*Plpe Segmont &tween Acc.
~
28 RPV,Nozzlo Forging OUtlOI 10-13 Discharge Header & RCS Isolation Valves 15 16 AFW-Maln Steam to AFW Pump Turbine Drive
. RCS*Pfpe Segment Between Loop Stop Valve 29 AFW-AFW TOP SUctlon Una LPI-Plpe Segment Between Containment & RPV (Cokl Log) 30 AFW-Plpe Segment from Emergency Isolation Valvo (lnslde) & Cold Leg Injection 17 LPI-LPI Pu"lJ SUctlon Une Makeup System, Fire Main LPl*Plpe Segment Between Containment !sol. 18 RCS-Pressurizer Spray Line 31 RCS-Pressurizer ReMef/Safat)'U Una Valve (Inside) & Hot leg Injection 19 RCS*Plpe Segment Betwoen RPV & Loop Stop 32 RCS-Pre&surizer surge Una 9 LPl*Sourco (RWST,Sump), Supply Lino Volvo(Hot Log) 33 LPl*A* Aca,uwlator Discharge Une 1o-14 10 LPl*Plpe Segment B9tween Pump Discharge &
Containment lsol. Valve 20 21 AFW*AFW TD Pump Otscharge Une AFW-Pipe Segment from Unit AFW PL1f11JS 34 35 RCS-Pipe Segment Betwaen SG and RCP RCS-Pipe Segment Between Loop 36 RCS-Pipe Segmortt Bo1woon S!cp Valve & SO (Hot Leg) RCP & Loop Slop Volvo (Cold Log) 37 LPl*A*Ac:cumutator, Suction Uno 0 5 10 15 20 25 30 35 Component Identification R9111050.3 w b
Figure 3.6. Risk Contributions of Surry Components 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 LPI
- Pipe Segment.Between 23 RPV
- Nozzle to Vessel Welds 1.0 X 10*6 Containment !sol. Valve (Inside) and Hot Leg lnJectlon 24 25 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
- 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 37 101 10° 10-1
>, 10-2 e0 Cu Cl) 0 :::s ca CT 10-3 C l!!
ou.
Ea, "C C> 10-4 C ca o
oEca Q Rank Syatem Component 10-5 1 RPV*B1ttllm Region-Welda
,a 20 RPV.Jnatrument Unn RPV*Vnul stucb 3
RPV-8.tlllne Plate RPV*Lower/Bottom Shell 21 22 RC&Pr. .1urlztrlR1tlef Safety Une RCS-Preasurtzor Surge Un, 4 RPY-Clr. Flange to Nozzle Course, Upper Shell. OUtalde Bettllne Wolds 23 RCS-Pipe Segment Between SO and ACP 5 LPI-Soun:a (RWST, Sump~ Supply Uno 24 RCS-Pipe Sogmonl Bolwnn Loop Slop Volvo & SO (Hot Log) 10-s 8 7
LPI-Plpe s.an,ent BetwNn Containment laol. v,rv. (ln11d1J
- Cold Ltg lnjectlon LPI-Plpe Segment Between Pump DlKharge & Containment lso. Valve 25 26 RCS-Pipe Segment Botwnn RCP & Loop Slop Volvo (Cold Log)
AFW*Plpo Segment Betwwn Containment Isolation Valve and SQ lsollltlon Yalvn 8 LPI-A*Pfpe Segment Betwnn Acc. Dlac:harge Headtr & RCS lsolaUon ValvH 27 AFW*ltolatlon Valvo to SO 9 AFW-CST, Supply Uno . 28 AFW-AFW TD Pump Dlschfllrgo Une 10 LPI-Plpe Segment Between Containment laol. Valve (lnald1) & Cold Leg ln)ectlon 29 AFW-lllln Stum to AFW Pump Turbine Drlvo 11 LPI-Plpe Sogment Between Containment lsoL V1lvo1. 30 AFW*Plpe Segment from UnH MW Pumpt 12 RCS-Plpo Segment lletwHn Loop Slop Volvo & RPV (Cold Leg) 31 RCS-Preasurtzer Spray Uno 10-1 13 14 RPV.ffozzl1 to Venell Welda RCS-Plpo Segment lletwHn RPV & Loop Slap Volvo (Hot Leg) 32 33 AFW*AFWTDP Suction Uno AFW*AFW UDP Dl,chorgo Uno 15 RPV-Uppor era...., Hood Flange 34 AFW*AFW UDP Suction Uno 18 ~LPI Pump Section Una 35 LPI-A*Accumulalor Discharge Une 17 RPV-Nozzlo Forging lnlol/Oullot 38 AFW*Plpo Segment from Emergency Makeup 5yalem, Flre Main 18 RPV-CRDMS :n LPI-A*Accumulalar, Suction Une 10-a 0 5 10 15 20 25 30 35 40 Component Identification w
b R9111050.4 Figure 3.8. Risk Contribution of Surry Components Based on Conditional Core Damage Given the Rupture
4.0 Discussions of the Results This section discusses the results presented in Sec- these components would disable the entire AFW sys-tion 3.0. Proposed values of target risk and rupture tem, and thus contribute significantly to core damage.
probability are also discussed. The discussions are based on the estimated median parameter values. Relatively high rankings were estimated for the pipe segments of the LPI-Accumulator system extending 4.1 Ranking of Component Risk from the accumulator discharge headers to the RCS isolation valves. The high rankings of these lines are The rankings of Table 3.1 were developed on the basis due to their important safety functions in providing of core damage frequency. In this discussion we will coolant to the RCS following an accident. Because the identify. the factors that govern these rankings, begin- residual heat removal (RHR) system is connected to ning with the highest ranked components and ending these lines, ruptures of these lines could prevent cooling with the lowest ranked components. water from being supplied to the RCS loops during the plant shutdown or cooldown.
For discussion purposes, "high-risk importance compo-nents" refer to components that have core damage Pipe segments of the LPI system extending from the frequency between l.OOE-08 and > l.OOE-06. Similarly, inside containment isolation valves to the RCS cold-
"medium-risk importance components" refer to those and hot-leg injection headers were also identified to be components having core damage frequencies of 1.00E- high risk-important components. The high rankings are 10 to l.OOE-08. "Low-risk importance components" due to relatively high stresses, potential for overpres-refer to those components having core damage frequen- surization of these lines, and the important functions of cies less than l.OOE-10. these lines in providing coolant to the RCS following a large LOCA. A high risk importance is also noted for 4.1.1 High-Risk Importance Components 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 he most risk-important components are located within LPI pump discharge lines up to the containment isola-the beltline region of the reactor vessel. The impor-tion valves follow in risk importance due to either lower tance of this region is due to high neutron fluence and stresses and/ or lower core damage consequences result-associated embrittlement levels, in conjunction with the ing from component ruptures. Depending on the break high consequence ( core damage) resulting from struc-location, ruptures of these lines result in an interfacing tural failure. The critical welds were identified on the system LOCA outside the containment.
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 4.1.2 Medium-Risk Importance Components is certain. The lower and bottom shell portions of the reactor vessel were also important but had somewhat The control rod drive mechanisms (CRDMs) and in-lower rankings due to the lower estimated rupture strument lines of the reactor vessel are next in ranking.
probabilities of these components ( e.g., lower ftuence These components have much lower estimated conse-and embrittlement levels). The circumferential flange- quences of failures compared with other RPV compo-to-nozzle course welds, upper shell welds, and welds nents. The failure consequences were estimated in the outside the beltline region of the pressure vessel had worst case to be equivalent to a large LOCA. Equal still lower rankings but were identified to be important. importance was estimated for the pipe segments of the Although these components have lower neutron fluence AFW system extending from the containment isolation and lower enibrittlement levels, failures of these compo- valves to the steam generator isolation valves, and also nents also have high core damage consequences. the pipe segments extending from steam supply lines to the AFW pump turbine drive. The importance of these High-risk importance (approaching RPV components) lines is due to a combination of high stress and high
- for the AFW system supply lines and sources (e.g., system unavailability resulting from a line rupture.
condensate storage tanks) was due to relatively active failure mechanisms (e.g., corrosion). Although these
.components have relatively low pressures, failure of 4.1 NUREG /CR-6181
4.0 Discussions As shown in Table 3.1, the next most important seg- were estimated in the worst case to be equivalent to a ments are the main RCS piping from the cold-leg loop large LOCA. The AFW motor-driven pump suction stop valves to the pressure vessel, the pressurizer spray and discharge lines, the turbine-driven pump suction piping, and the pipe segments from the pressure vessel line, the pipe segments from the emergency makeup to the hot-leg loop stop valves. Failure of any of these system, and the fire main were estimated to have rela-lines results in a large LOCA which cannot be isolated tively low rankings. This is due mainly to the low esti-by the loop stop values, as is the case with other seg- mated consequences for failures of these components.
ments of the main RCS piping. High estimated con-sequences resulting from lines being connected to the Of the 37 components ranked in Table 3.1, the follow-RCS loop (e.g., safety injection, RHR lines, etc.) in- ing were identified to have the lowest importance: the crease the importance of these particular pipe segments pressurizer relief and safety lines, pressurizer surge line, within the reactor coolant loop. The importance of the pipe segments extending from the hot-leg loop stop pressurizer spray line results from a relatively high- valve to the steam generator, pipe segments extending estimated failure probability due to thermal stresses and from the coolant pump to the cold-leg loop stop valve the key function of this line in controlling the desired of the RCS, and the accumulator discharge and suction primary system pressure. Failure of the spray line lines. These low rankings are due to low rupture prob-could result in LOCAs in Loops A and C, in addition ability estimates and/or low core damage consequence to the loss of the pressurizer function. estimates for these components. Additionally, failures of the lines within the RCS loops, at most, will result in Of equal importance in Table 3.1 are the pump suction a large LOCA which can be isolated by the loop stop lines of the LPI system, the AFW turbine-driven pump values.
discharge lines, the Unit 2 AFW pump cross-connected line, and the pipe segments extending from the AFW The cumulative risk contribution for all 37 components
- isolation valves to the steam generators. Ruptures of of the four systems ( as shown in Figure 3.6) is about the LPI pump suction lines would prevent the borated 2.lE-06 per plant year. Significant contributions to risk water from being supplied to the RCS when needed. come only from failures of approximately the first Similarly, a break in the AFW turbine-driven pump 10 components. The welds of the beltline region of the discharge line could prevent sufficient cooling water reactor pressure vessel dominate the risk, accounting from being supplied to the steam generators. Unless for almost 75% of the core damage frequency due to appropriate recovery actions are taken, a complete loss component failures. The beltline welds are followed in of water supply to the steam generators may result, importance by the beltline plate material, which contributing significantly to core damage. The impor- accounts for another 5%. The welds in the upper and tance of the Unit 2 AFW pump cross-connected line is lower reactor heads account for another 6%; the single due to its key function in providing cooling to the steam AFW condensate storage tank and supply line contrib-generators in the case that Unit 1 AFW is lost. This ute 3%; and various welds in the LPI system contribute cross-connected line is used for mitigating other initi- another 10%). This adds up to more than 99% of the.
ating events as well (e.g., station blackout). Finally,
- total core damage frequency risk associated with com-failures of the pipe segments extending from the AFW ponent ruptures for the four system analyzed. The isolation valves to the steam generators would result in system level rankings derived from component contribu-steam generator blowdown through the break (similar tions to core damage are the following: 1) RPV, 2) to a main steam line break) and a loss of secondary LPI, 3) AFW, and 4) RCS. These results agree with cooling. those obtained from the earlier PNL system-level study (Vo et al. 1989).
4.1.3 Low-Risk Importance Components Table 4.1 presents the Surry-1 plant-specific ASME Lower importances are noted for the RPV nozzle to classifications and required ISI examinations for each vessel welds, vessel studs, the RPV upper closure head piping section or component of Table 3.1. Table 4.1 and flanges, and the RPV nozzle forging (inlet and shows that ASME classifications and ISi requirements outlet). The failure consequences of these components are in general agreement with the importance rankings based on cor~ damage frequency. In particular, the
'.,I.,,'
NUREG/CR-6181 4.2
4.0 Discussions RPV component making the greatest contribution to Examination of Table 3.1 shows that the total risk for the core damage frequency has the most stringent in- the four systems analyzed is about 2.lE-06 per plant spection requirements. However, recommendations for
- year, which is rather close to the total recommended setting inspection requirements based solely on Table target risk of 2.5E-06 per plant year. Thus, the total
- 4.1 should be made cautiously because additional plant target risk in this case is essentially the same as the systems will be considered in future PNL work. total estimated risk due to ruptures. However, analysis of the remaining systems at Surry-1 may increase this 4.2 Development of Target Risk and total risk until it may equal or exceed the total recom-mended target risk.
Rupture Probability Values The recommendation that the target risk be appor-A philosophy and approach for selecting target risk tioned on the basis of the component's estimated risk, values and target rupture probabilities has been recom- in this case, means that the target risk values can be set
- mended by the ASME Research Task Force on Risk- equal to the component's estimated risk. This, in turn, Based Inspection. The philosophy is that the inspection means that the target rupture probability for each com-should ensure that the risk of core damage resulting ponent can be the same as the rupture probability from pressure boundary and structural failures should estimated by the expert judgment elicitation panel be a small fraction of the total core damage risk esti- (since consequences are fixed by the PRA analysis given mated in the plant specific PRAs. The ranking process that a rupture occurs). Table 3.3 of Section 3.0 shows described in this report can be used to set priorities for the rupture probabilities for components of Table 3.1.
inspection but does not provide criteria for determining For example, the beltline region of the Surry-1 RPV the degree of inspection. For this purpose, the risk due contains five welds (four longitudinal welds and one to pressure boundary and structural failures is herein circumferential weld). The per-weld rupture probability eferred to as the "target risk," and 5% of the total was estimated as 3.16E-07, and the estimated target risk RA estimated risk from internal events has been for all five welds was 1.58E-06 (Table 3.1). Based on recommended as an appropriate value. It is further the above discussions, the value of 3.17E-07 can be used recommended that this target risk be apportioned un- as a desired target rupture probability for each weld equally among the risk-important components by con- within the RPV beltline region.
sidering the estimated risk associated with rupture of each component (Gore, et al. 1991). Using the results Following this approach, the objective of ISi is to pro-of conditional probability of core damage given compo- vide confidence that the failure probabilities 'do remain nent failures, then, the target rupture probabilities for . at or below the values estimated by the expert elicita-components can be estimated. tion. In those cases where the total risk for all com-ponents exceeds target values ( e.g., 2.5E-06 per plant The core damage frequency estimated in state-of-the-art year), and an additional objective of ISi should be the PRAs for modern facilities is about 5.0E-05 per plant reduction of failure probabilities for selected compo-year. Using this number, an appropriate target risk nents.
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 NUREG /CR-6181
4.0 Discussions 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 RPV Beltline Welds B-A Volumetric RPV Beltline Plate B-A Volumetric RPV Lower /Bottom Shell B-A Volumetric AFW CST, Supply Line D-B Visual RPV Circumferential Flange to Nozzle Course, Upper B-A Volumetric Shell, Outside Beltline Welds LPI-A Pipe Segment Between Accumulator Discharge B-I Volumetric and Header and RCS Isolation Valves Surface LPI Pipe Segment Between Containment Isolation Valve B-J, C-F-1 Volumetric and (inside) and Cold-Leg Injection Surface LPI Pipe Segment Between Containment Isolation Valve B-J, C-F-1 Volumetric and (inside) and Hot-Leg Injection Surface LPI LPI Sources (RWST, Sump), Supply Line D-C Visual LPI Pipe Segment Between Pump Discharge and Con- C-F-1 Volumetric and tainment Isolation Valve Visual LPI Pipe Segment Between Containment Isola~ion Valves C-F-1 Volumetric and Visual RPV CRDMs B-E Visual RPV Instrument Lines B-J Exempt AFW Pipe Segment Between Containment Isolation and C-F-1 Volumetric and SG Isolation Valves Surface AFW Main Steam to AFW Pump Turbine Drive C-F-1, Volumetric, RCS Pipe Segment Between Loop Stop Valve and RPV B-J Volumetric (Cold Leg)
LPI LPI Pump Suction Line C-F-1 Volumetric and Surface RCS Pressurizer Spray Line B-J Volumetric RCS 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 Risk-based priorities were compared with the current and has been applied in a pilot study to identify and . Surry-1 plant-specific ASME classifications and required prioritize the most risk-important systems and compo- ISi examinations. The ASME classifications and ISi nents at Surry-1. In the pilot application, the method requirements are in general qualitative agreement with used component failure probabilities estimated from the risk-based rankings based on core damage frequency.
expert judgment elicitation conducted by PNL, results The components making the greatest contributions to from PNL's Surry-1 system prioritization, and system the core damage frequency currently have the most fault trees reported in the Surry-1 PRA to prioritize to stringent inspection requirements. However, final con-the high-priority components for inspection. clusions for setting inspection requirements should await further pilot studies.
As shown in Table 3.1, contributions of component failures to core damage frequency range widely from An approach for determining target risk and target about l.6E-06 to 1.6E-14 per plant year. The cumula- rupture probability values has been proposed and sue~
tive risk contribution (as shown in Figure 3.5) is about cessfully pilot tested for components within the selected 2.lE-06 per plant year. This estimate is about 5% of systems at Surry-1. It is recognized that in some cases the total Surry-1 PRA risk. The total estimated risk is the estimated target values of rupture probabilities may dominated by failures of the reactor pressure compon- be difficult to achieve, therefore, further studies are ents (86%). This risk is followed by the LPI system n~ded to determine whether this approach is generally components (10% ), and then other components within
- appropriate.
the AFW and RCS (4%). The results provide a guide to establish improved inspection priorities for nuclear The analysis for the Surry-1 plant will be completed by power plant components. developing the risk importance of components in the remaining systems (e.g., high-pressure injection, service sitivity analyses addressed uncertainties on parame- water, and balance of the plant). Similar analyses will values and modeling assumptions. The sensitivity of be performed for other PWRs and boiling-water reac-mponent rankings to upper- and lower-bounding . tors (BWRs), and generic trends in component impor-values of estimated rupture probabilities was estab- tances will be established. Once the high-priority com-lished. As shown in Tables 3.1 and 3.2, the results ponents have been identified, recommended inspection indicated no significant changes in component rankings. programs (method, frequency and extent) will be devel-Additional sensitivity analyses addressed contributions oped. Probabilistic structural mechanics and decision to core damage frequency due to indirect effects of analyses will be applied to establish inspection strategies component failures. The results indicate that the over- that will ensure that component failure rates are main-all contribution to core damage frequency from the tained at acceptable levels and in a cost-effective man-indirect effects was negligible. Sensitivity and uncer- ner.
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.
- 5.1 NUREG /CR-6181
6.0 References
~ Resea<ch Task Force on llisk-Based Inspection Guidelines. 1991. Risk-Based Inspection - Development U.S. Nuclear Regulatory Commission (NRC). 1989.
Severe Accident Risks: An Assessment for Five U.S.
of Guidelines, Volume 1 General Document. CRTD- Nuclear Power Plants. NUREG-1150, Summary Report, Vol, 20-1, American Society of Mechanical Engineers Second Draft For Peer Review, U.S. Nucl~ar Regulato-Center for Research and Technology Development. ry Commission, Washington, D.C.
Bertucio, R. C., and J. A. Julius. 1990. Analysis of Vo, T. V., B, F. Gore, L. J. Eschbach, and F. A.
Core Damage Frequency: Surry, Unit 1 Internal Events. Simonen. 1989. "Probabilistic Risk Assessment-Based NUREG/CR-4550, Sandia National Laboratories, Albu- Guidance for Piping Inservice Inspection." Nuclear querque, New Mexico.
- Technology, Volume 88 (1), American Nuclear Society, La Grange Park, Illinois.
Gore, *B. F., T. V. Vo, and K. R. Balkey. 1991. "Status of ASME Rick-Based Inspection Guidelines Develop- Vo, T: V., B. W. Smith, F. A. Simonen, and S. R. Doc-ment for Nuclear Power Plants." To be presented at tor. 1990. "Development of Generic Inservice Inspec-the 1991 ASME Winter Annual Meeting, Atlanta, . tion Guidance for Pressure Boundary Systems. Nucle-Georgia.
- ar Technology, Volume 92 (3), American Nuclear Soci-ety, La Grange Park, Illinois.
Heinecke, C. C., V. A. Perone, M. Weaver, and C. N:
Wright. 1987. Surry Units 1 and 2 Reactor Vessel Flu- Vo, T. V., P. G. Heasler, S. R. Doctor, F. A. Simonen, ence and RTPTS Electrons. WCAP-11015, Westinghouse and B. F. Gore. 1991. "Estimate of Component Rup-Electric Corporation, Pittsburgh, Pennsylvania. ture Probabilities. Expert Judgment Elicitation." Nu-clear Technology, Volume 94 (1), American Nuclear Meyer, M.A., and J.M. Booker. 1989. Eliciting and Society, La Grange Park, Illinois.
Analyzing &pert Judgment. NUREG/CR-5424, Los os National Laboratory, Los Alamos, New Mexico. Wheeler, T. A., S. C. Hora, W.R. Cramond, and S. D.
Unwin. 1989. Analysis of Core Damage Frequency from ussel, K. D., et al. 1987. Integrated Reliability and Internal Events: &pert Judgment Elicitation.
Risk Analysis System (IRRAS). NUREG/CR-4844, NUREG/CR-4550, Volume 2, Sandia National Labora-Idaho National Engineering Laboratory, Idaho Falls, tories, Albuquerque, New Mexico.
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 NUREG/CR-6181
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.
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 1 OF 13 PROGRAM: NOE-IMPROVED IS!
SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: FLANGE TO NOZZLE/UPPER SHELL WELD (1) (2) (3) (4) (5) (6) (7)
Failure Component Probabi lit) a) Core Damffi ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Weld 1 - circumfer- ( 157" ID Core damage was No recovery 2.00E-08 4
- Low fluence and low ent i al flange to 9" thick) assumed embrittlement nozzle course
- The RPV was assumed to 3.16E-08 p cm/s = 1. 0 fail given a break at this 2.00E-08 location 1.00E-08 P/Pf = 1. 0 CDP= 1. 0 Weld 2 - circumfer- (157" ID Core damage was No recovery 3 .16E-08 3
- Lower fluence and lower ential upper shell 9" thick) assumed embrittlement to weld, intermedi-
- Cu content= 0.11%
ate shell 1. OE-07 p cm/s = 1. 0
- Ni content= 0.7%?
3.16E-08
- 1. OE-08 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 Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) p cm = Pf*~ p cmI s 1 *Ps1 I Pf*R jt 1I (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 .
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 2 DF 13 PROGRAM:
SYSTEM:
NDE-IMPROVED !SI REACTOR PRESSURE VESSEL COMPONENT: LOWER SHELL/BOTTOM HEAD (1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Damffi ImportfBJe (location) (component size) Fai 1ure Effect Recovery Action Frequency Index Remarks Weld 3 - lower (157" ID Core damage was No recovery 3.16E-07 1
- High fluence and high shell 9" thick) assumed embrittlement. High ther-mal stress for LOCA and
- 1. OE-06 p = PTSs.
cm/s 1. 0 3.16E-07
- Critical weld based on
- 1. OE-07 P/Pf = 1.0 Surry PTS study (WCAP-11015, 1987).
CDP = 1.0
- Cu content= 0.11%
- Ni content= 0.7%?
Weld 4 - bottom (157" ID Core damage was No recovery 3.16E-08 3
- Lower fluence and lower head 5.4" thick) assumed embrittlement
- Cu content= 0.11%
- 1. OE-07 pcm/s = 1. 0
- Ni content= 0.7%?
3.16E-08 5.2E-09 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 Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(cl pcm= Pf*~ pcm/s1*PsJ Pt'*R1I
~
(d) Based on "Median Values" of failure probabilities.
tI1 NOTE: Q - never used on these tables.
C) CDP - core damage probability given the break.
......... EDP - emergency operating procedure.
Q LOCA - loss of coolant accident.
O'I PTS - pressurized thermal shock.
~
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET- 3 OF 13 PROGRAM: NOE-IMPROVED ISi SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: BOTTOM HEAD/INTERMEDIATE SHELL
( 1) (2) (3) (4) (5) (6) (7)
Fai 1ure Component Probability(a) Core Dam'.Y:m ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Weld 5 - bottom ( 157" ID Core damage was No recovery 3.16E-08 3
- Low fluence and low head 5.4" thick) assumed thermal stress.
- 1. E-07 p -
cm/s - 1. 0 3.16E-08
- 5. 2E-09 P/Pf = 1. 0 CDP = 1.0 Weld 6 - longitudi- (100" 1ength) Core damage was No recovery 3.16E-07 1
- High fluence and embrit-nal weld intermedi- assumed tlement at lower end of ate shell 5E-07 this weld. High thermal 3.16E-07 p stress for LOCA and PTS.
- l. lE-07 cm/s = 1. 0
- Cu content= 0.11%
P/Pf = 1.0
- Ni content= 0.7%?
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.
(bJ 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.
- FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 4 OF 13 PROGRAM: NOE-IMPROVED ISI -.
SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: INTERMEDIATE SHELL (1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Dam'.'{ir) ImportfBJe
. (location) (component size) Failure Effect Recovery Action Freauency Index Remarks Weld 7 - longitudi- (100" length) Core damage. was No recovery 3.16E-07 1
- High fluence and embrit-nal weld interme- assumed tlement at lower end of di ate she 11 5.0E-07 this weld. High thermal p = stress for LOCA and PTS.
3.16E-07
. cm/s 1. 0
- Cu content= 0.11%
- l. lE-07 P/Pf = 1.0
- Ni content= 0.7%?
CDP = 1. 0 Weld 8 - longitudi- (100" length) Core damage was No recovery 3.16E-07 1
- High fluence and embrit-nal weld lower assumed tlement at upper end of shell 1. OE-06 this weld. High thermal 3.16E-07 p stress for LOCA and PTS.
cm/s = 1. 0
- Critical weld based on
- l. lE-07 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 Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) Pcm= Pf*~ Pcm/s 1 *Ps/ Pf*R1I (d) Based on "Medi an Values" of fa.il ure probabi 1it i es.
~tI1 NOTE: Q- never used on these tables.
0 CDP - core damage probability given the break.
......... EDP - emergency operating procedure .
~ LOCA -
PTS -
loss of coolant accident.
pressurized thermal shock.
....O'I 00
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 5 OF 13 PROGRAM: NOE-IMPROVED !SI SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: INTERMEDIATE SHELL/NOZZLE TO VESSEL WELD (1) (2) (3) (4) (5) (6) (7)
Failure Component Probabil i ty(a) Core Damffi ImportfBJe (location) (comoonent size) Failure Effect Recovery Action Frequency Index Remarks Weld 9 - longitudi- (100" length) Core damage was No recovery 3.16E-07 1
- High fluence and embrit-nal weld lower assumed tlement at upper end of shell 1. OE-06 this weld. High thermal cm/s = 1. a 3.16E-07 p stress for LOCA and PTS.
l.lE-07
- 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%?
Weld 10 - nozzle- (app. 40" dia.) L/LOC,A (see re- Follow EOPs 1. OOE-11 7
- Lower fluence and low to-vessel weld, mark) embrittlement.
Loop 1 outlet 3.16E-08
- Break at this location 2.0E-08 pcm/s = 5.00E-04 was assumed to be equiva-
- 1. lE-08 lent to a large LOCA.
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.
(b) Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) Pcm= P.t'*l'f' Pcm/s *P8 rt 1 , 1
/ P.t'*RJ
~
(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.
- FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 6 OF 13 PROGRAM: NOE-IMPROVED ISI SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: NOZZLE TO VESSEL WELDS (1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Dam'.¥1:n ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Weld 11 - nozzle- (app. 40" dia.) L/LOCA was No recovery 1. OOE-11 10
- Lower fluence and low to-vessel weld, assumed given a embrittlement.
Loop 1 inlet 3.16E-08 break 2.0E-08
- 1. OE-08 Pcm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 '
Weld 12 - nozzle- (app. 40" dia.) L/LOCA was as- Follow EOPs 1. OOE-11 10
- Lower fluence and low to-vessel welds, sumed given a embrittlement.
Loop 2 outlet 3.16E-08 break 2.0E-08
- 1. OE-08 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 Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(cl p= = Pf*~ pcm/s1 *Ps1 f Pf*Ri' (d) Based on "Median Values" of failure probabilities.
~
~ NOTE: Q - never used on these tables.
CDP - core damage probability given the break.
0
......... EOP - emergency operating procedure .
~ LOCA PTS loss of coolant accident.
- pressurized thermal shock.
O'I I-'
00 I-'
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 7 OF 13 PROGRAM: NOE-IMPROVED IS!
SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: NOZZLE TO VESSEL WELDS
( 1) (2) (3) (4) (5) (6) (7)
Failure Component Probabi 1ity (a) Core Damffi ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Weld 13 - nozzle- (app. 40" di a.) L/LOCA was Foll ow EOPs l.OOE-11 10
- Lower fluence and low to-vessel weld, assumed given a embrittlement.
Loop 2 inlet 3.16E-08 break 2.0E-08 l.OE-08 pcm/s = 5.00E-04 P/Pf = l. 0 CDP = 5.00E-04 Weld 14 - nozzle- (app. 40" dia.) L/LOCA was as- Fo 11 ow EOPs l.OOE-11 10
- Lower fluence and low
- to-vessel welds, sumed given a embrittlement.
Loop 3 outlet 3.16E-08 break 2.0E-08
- l. OE-08 p = 5.00E-04 cm/s 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/s 1 *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 .
PROGRAM: NOE-IMPROVED ISI FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 8 OF 13
- SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: NOZZLE TO VESSEL WELD/NOZZLE FORGINGS (1) (2) (3) (4) (5) (6) (7)
Fai 1ure Component Probabi 1it)al Core Dam'.Y:tr) ImportfBJe (location) (component size) Failure Effect Recovery Action Freauency Index Remarks Weld 15 - nozzle- (app. 40" dia.) L/LOCA was Follow EOPs 1. OOE-11 10
- Lower fluence and low to-vessel weld, assumed given a embrittlement.
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 10
- High stress at nozzle inlets for Loops 1, sumed given a corner. Potent i a1 for 2, and 3 5.0E-08 break high thermal stress from 3.0E-08 cold fluid injection.
- 1. OE-08 pcm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 (a) Component fai 1ure probabi 1i ty obtained from Vo et al . 1991. The three va 1ues repres*ent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl Importance measure or rank relative to other components of the RPV, based on "Median Va 1ues" of failure probabilities.
(c) Pcm = Pt*~ pcm/s 1 *Ps,f Pt*R1I (d) Based on "Median Values" of failure probabilities.
~~
NOTE: Q - never used on these tables.
-°'
Q t'I1 Cl I-'
00 CDP EDP LOCA PTS core damage probability given the break.
emergency operating procedure.
loss of coolant accident.
pressurized thermal shock.
I-'
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 9 OF 13 PROGRAM: NOE-IMPROVED ISi SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: NOZZLE FDRGINGS/BELTLJNE PLATE (1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Dam'.¥tr) ImportfBJe (location) (component size) Fai 1ure Effect Recovery Action Frequency Index Remarks Nozzle Forgings - (app. 40" dia.) L/LOCA was Follow EOPs 1. DDE-11 10
- High stress at nozzle outlets for Loops assumed given a corner.
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 2
- High fluence, but lower (base metal) 9" thick) assumed embrittlement rate than welds.
p
- 1. DE-07 cm/s -- 1. D
- 1. DE-07 8.DE-08 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.
(b) Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) p cm = P.)Y' L jt p cmI s 1 *Ps 1 I P~*R*I L i (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.
PROGRAM:
SYSTEM:
NOE-IMPROVED ISI REACTOR PRESSURE VESSEL FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 10 OF 13 COMPONENT: VESSEL SHELL/UPPER HEAD *.
(1) (2) (3) (4) (5) (6) (7)
Fai 1ure Component Probability(a) Core Damffi ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Vessel Shell (157" OD Core damage was as- No recovery 1.0DE-08 5
- Lower fluence and less outside beltline 9" thick) sumed given a break at embrittlement.
(base metal) this location 3.16E-08
- 1. OE-08 pcm/s = 1. O
- Low fluence and low metal) 6" thick) given a break at this embrittlement.
location 5.DE-08 3.16E-08 pcm/s = 5.00E-04
- 1. OE-08 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 Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) p cm = Pf*IY" rt p cmI s 1 *Ps 1 I Pf*R1I Based on "Median Values" of failure probabilities.
~~
(d)
NOTE: Q- never used on these tables.
-°'
t'I1 0
~
I-'
00 CDP -
EDP -
LOCA -
PTS -
core damage probability given the break.
emergency operating procedure.
loss of coolant accident.
pressurized thermal shock.
I-'
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 11 OF 13 PROGRAM: NOE-IMPROVED !SI SYSTEM: REACTOR PRESSURE VESSEL COMPONENT: LOWER HEAD/VESSEL FLANGE
( 1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Damffi ImporttBre (location) (component size) Failure Effect Recovery Action Freauencv Index Remarks Lower Head (base (157" OD Core damage was No recovery 1. ODE-08 5
- Lower fluence and low metal) 5.4" thick) assumed embrittlement.
cm/s = 1. a 5.DE-08 p
- 1. OE-08
- 1. OE-08 P/Pf = 1.0 CDP = 1. 0 Vessel Flange ( 149" ID x 184" L/LOCA was as- No recovery l.OOE-11 10
- Low fl uence. Stress OD) sumed given the concentr~tion from closure break studs.
5.0E-08
- Potential surface imper-8.0E-08 p fection. However, it has cm/s = 5.00E-04 been eliminated; there-
- 1. DE-08 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.
j NOTE: Q- never used on these tables.
COP - core damage probability given the break.
EDP -
LOCA -
emergency operating procedure.
loss of coolant accident. .*
PTS - pressurized thermal shock.
- PROGRAM:
SYSTEM:
COMPONENT:
NOE-IMPROVED !SI REACTOR PRESSURE VESSEL VESSEL FLANGE/STUDS FAILURE MODES AND EFFE ALYSIS WORKSHEET SHEET 12 OF 13 (1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Damffi ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Closure Head (149" ID x 184" L/LOCA was as- No recovery l.OOE-11 10
- Low fl uence. Stress concen-Flange OD) sumed given a tration from closure studs.
break at this
- Potential leakage at RPV 5.0E-08 location flanged leak off line (1" 2.0E-08 line) due to 0-ring or valves
- 1. OE-08 p (upstream) leakage.
cm/s = 5.00E-04 P/Pf = 1.0 CDP= 5.00E-04 Reactor Vessel (app. 6" dia. L/LOCA was as- No recovery 5.00E-11 8
- Multiple failure of vessel Studs (58 studs) each) sumed given the studs must be postulated for break (see re- LOCA.
- 1. OE-06 marks)
- Some stud nicks -and gauge
- 1. OE-07 problems.
3.16E-08 p
- Multiple stud failures con-cm/s = 5.00E-04 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) Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) Pcm = Pt*~ Pcm/s 1 *Ps1 / Pt*Ril
~~ (d) Based on "Median Values" of failure probabilities.
tI1 NOTE: Q - never used on these tables.
0
.......... CDP - core damage probability given the break .
("") EDP - emergency operating procedure.
~ LOCA - loss of coolant accident.
0\ PTS - pressurized thermal shock.
00
~
-~
FAILURE MODES AND EFFECTS ANALYSIS WORKSHEET SHEET 13 OF 13 0
PROGRAM: NOE-IMPROVED ISi I-' SYSTEM: REACTOR PRESSURE VESSEL 00 I-' COMPONENT: CRDNs/INSTRUMENT LINES (1) (2) (3) (4) (5) (6) (7)
Failure Component Probability(a) Core Damffi ImportfBJe (location) (component size) Failure Effect Recovery Action Frequency Index Remarks Control Rod Drive (app. 4" dia.) L/LOCA was as- No recovery 5.DOE-09 6
- Contribution from multiple Mechanism (~5 sumed given a CROM failures were insignifi-penetrations of 5.48E-05 break at (see cant (e.g., using Chi-square upper head) 1. OE-05 remarks) distribution to estimate the 4.47E-06 failure effect due to multiple p CROM fail ures) .
cm/s = 5.00E-04 P/Pf = 1. D CDP= 5.DOE-04 Instrument Line (app. 1" di a.) L/LOCA was as- No recovery 5.DDE-09 6
- Cracking and repairs have Penetrations sumed been reported at some plants.
(lower head, 1o'o 8.37E-05
- Seal table (thimble tube) instrument lines) 1. DE-05 p cm/s = 5.0DE-04 leakage due to seal fitting 3.16E-06 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.
Lower Head Skirt -- -- -- -- --
- Laying on concrete support.
(a) Component failure probabflity obtained from Vo et al. 1991. The three values represent the median, and 25% and 75% quartiles of uncertainty distribution.
(bl Importance measure or rank relative to other components of the RPV, based on "Median Values" of failure probabilities.
(c) p cm = rt Pf*IY' p cmI s 1 *Ps 1 I Pf*R1I (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 -
PTS -
loss of coolant accident.
pressurized thermal shock.
.~
NUREG-CR/6181 PNL-9020 RS DISTRIBUTION No. of No. of Copies Copies OFFSITE FOREIGN 2 A. J. Hiser, Jr. J. R. Tomlinson NRC/RES NDT Application Centre Mail Stop NS 217C Nuclear Electric plc Timpson Road M.R.Hum Wythenshawe NRC/NRR Manchester M23 9LL Mail Stop 7 D4 United Kingdom G. Johnson ONSITE NRC/RES Mail Stop 7 D4 50 Pacific Northwest Laboratory J. Muscara E. S. Andersen NRC/RES R. E. Bowey Mail Stop NS 217C D. M. Boyd S. H. Bush D. W. Craig A. A. Diaz NRC/RES S. R. Doctor (18)
Mail Stop NS 217C J. R. Friley B. F. Gore (5)
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NRC/Region I J. C. Spanner T. V. Vo (5)
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.. ,I',- :>
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NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I 1. REPORT NUMBER 12-891 IAaitMd t,y NRC. Add Vol** $uclo., RIW.,
NRCM 1102, - ,.__'" Nummn, If ..v.l 3201, 3202 BIBLIOGRAPHIC DATA SHEET (SH instructions on rht1 rt1VtlrStlJ NUREG/CR-6181
- 2. TITLE ANO SUBTITLE PNL-9020 A Pilot Application of Risk-Based Methods 3. CATE REPORT PUBLISHED to Establish Inservice Inspection MONTH YEAR Priorities for Nuclear Components at AUCJUSt 1994 Surry Unit I Nuclear Power Station 4. FIN OR GRANT NUMBER 82289
- - 5. AUTHOR(Sl 6. TYPE OF REPORT T. Vo, 8. Gore, F. Simonen, S. Doctor 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,,.
,.,.;Jing _ _,
- -s.m. u *bo .."; ii canr-ror. pnwm NRC Dillirion. Offic* or R.,,ion, v.s. Nuc1.., RllflUl*rory CammiAion, 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 13. AVAILABILITY STATEMl:NT Unlimited nondestructive evaluation, probabilistic risk assessment, ASME 14. SECURITY CLASSIFICATION Code, inservice inspection, welds, piping systems, Inspection (This Page/
Importance, Birnbaum Importance, pressure boundary systems, Unclassified risk-based, Surry-1 (This Report/
Unclassified
- 15. NUMBER OF PAGES
- 16. PRIC.E NRC FORM 33S 12-l:19)