ML100560430
| ML100560430 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 02/23/2010 |
| From: | Price J Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 10-050 | |
| Download: ML100560430 (28) | |
Text
10 CFR 50.55a VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 February 23, 2010 U.S. Nuclear Regulatory Commission Serial No.10-050 Attention: Document Control Desk NL&OS/ETS RO Washington, D.C. 20555 Docket Nos. 50-338/339 License Nos. NPF-4/7 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
NORTH ANNA POWER STATION UNITS 1 AND 2 ASME SECTION XIINSERVICE INSPECTION PROGRAM RELIEF REQUEST N1-14-RI-001 AND N2-14-RI-001 REQUEST FOR ALTERNATIVE - IMPLEMENTATION OF A RISK-INFORMED INSERVICE INSPECTION PROGRAM BASED ON ASME CODE CASE N-716 Pursuant to 10 CFR 50.55a(a)(3)(i), Dominion hereby requests authorization to implement a risk-informed inservice inspection (RI lSI) program based on the American Society of Mechanical Engineers (ASME) Code Case N-716, as documented in the attached Request for Alternative N1-14-RI-001 and N2-14-RI-001. N1-14-RI-001 and N2-14-RI-001 are being submitted in a template format in Attachment 1. This template format is similar to the submittals the NRC Staff has approved for Waterford 3, Grand Gulf, and D. C. Cook 1 and 2. This format is also similar to the recently submitted request for alternative by Calvert Cliffs and Arkansas Nuclear One for the same subject.
In accordance with 10 CFR50.55a(a)(3)(i), the proposed alternative to the referenced requirements may be approved by the NRC provided an acceptable level of quality and safety are maintained. Dominion believes the proposed alternative meets this requirement.
Dominion requests to implement this alternative for the entire 4th lSI Interval for North Anna Units 1 and 2. North Anna Unit 14th 10 Year Interval began May 1, 2009 and will end April 30, 2019. North Anna Unit 24th 10 Year Interval begins December 14, 2010 and will end December 13, 2020.
Dominion requests review and approval of the attached relief requests by February, 2011 in order to plan and complete the first period examinations.
Serial No.10-050 Docket Nos. 50-338/339 Fourth Interval Risk Informed Relief Requests N1-14-RI-001 & N2-14-RI-001 Page 2 of 2 If you have any questions or require additional information, please contact Mr. Thomas Shaub at (804) 273-2763.
Respectfu lIy, rice sident - Nuclear Engineering Attachments
- 1. Request for Alternative - Relief Requests N1-14-RI-001 and N2-14-RI-001.
cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. J. E. Reasor, Jr.
Old Dominion Electric Cooperative Innsbrook Corporate Center 4201 Dominion Blvd.
Suite 300 Glen Allen, Virginia 23060 NRC Senior Resident Inspector North Anna Power Station Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 0-8 G9A 11555 Rockville Pike Rockville, Maryland 20852
Serial No.10-050 Fourth Interval RI Relief Requests N1-14-RI-001 and N2-14-RI-001 Page 1 of 26 Attachment INSERVICE INSPECTION PLAN RISK INFORMED FOURTH INTERVAL RI RELIEF REQUESTS N1-14-RI-001 AND N2-14-RI-001 North Anna Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)
Serial NO.1 0-050 Pg 2 of 26 REQUEST FOR ALTERNATIVE In accordance with 10 CFR SO.SSa(a)(3)(i)
DOMINION NORTH ANNA UNITS 1 AND 2 Relief Requests N1-14-RI-001 (Unit 1) and N2-14-RI-001 (Unit 2)
APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED / SAFETY-BASED mSERWCEmSPEcnON PROGRAM PLAN Table of Contents
- 1. Introduction 3 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 3 1.2 PRA Quality 3
- 2. Proposed Alternative to Current Inservice Inspection Programs 4 2.1 ASME Section XI 4 2.2 Augmented Programs 4
- 3. Risk-Informed / Safety-Based lSI Process 5 3.1 Safety Significance Determination 5 3.2 Failure Potential Assessment 5 3.3 Element and NDE Selection 8 3.3.1 Additional Examinations 9 3.3.2 Program Relief Requests 9 3.4 Risk Impact Assessment 10 3.4.1 Quantitative Analysis 10 3.4.2 Defense-in-Depth 12
- 4. Implementation and Monitoring Program 13
- 5. Proposed lSI Program Plan Change 13
- 6. References/Documentation 14
Serial No.10-050 Pg 3 of 26 NORTH ANNA POWER STATION UNITS 1 AND 2 REQUEST FOR ALTERNATIVE N1-14-RI-001 and N2-14-RI-001
- 1. INTRODUCTION North Anna Unit 1 (NAPS1) is currently in the fourth inservice inspection (lSI) interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section XI Code for Inspection Program B. NAPS1 4th Interval began May 1,2009. North Anna Unit 2 (NAPS2) begins the 4th Interval December 14, 2010. NAPS1 and NAPS2 plan to implement a risk-informed / safety-based inservice inspection (RIS_B) program during their entire fourth intervals.
The 4th Interval ASME Section XI code of record for both units at NAPS is the 2004 Edition for Examination Category B-F, B-J, C-F-1, and C-F-2 Class 1 and 2 piping components.
The objective of this submittal is to request the use of the RIS_B process for the inservice inspection of Class 1 and 2 piping. The RIS_B process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, which is founded in large part on the RI-ISI process as described in Electric Power Research Institute (EPRI)
Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure.
1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide (RG) 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, and RG 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.
1.2 Probabilistic Risk Assessment Quality The North Anna probabilistic risk assessment (PRA) model was originally developed in response to the NRC Generic Letter 88-20 on Individual Plant Examinations (lPE). The IPE was submitted to the NRC in December, 1992. This was accepted by the NRC and used for various applications. The PRA model and supporting documentation was maintained and updated to accurately reflect the current plant configuration and operating practices consistent with its intended application. In April, 2001, Dominion submitted a risk-informed inservice inspection (RI-ISI) program for North Anna to the NRC for approval. This program was developed in accordance with Westinghouse Owners Group WCAP-14572, Revision 1-NP-A, and was supported by the North Anna PRA model N7B which was released in March 1998. The NRC approved the program documented in a safety evaluation report in a letter dated September 18, 2001, stating the proposed RI-ISI program was an acceptable alternative to the requirements of ASME Boiler and Pressure Vessel Code,Section XI for inservice inspection of Class 1 piping.
The North Anna PRA model continued to be maintained and enhanced to accurately reflect the current plant configuration and operating practices. In 2000 the PRA model NOA was released. This model received a formal Westinghouse Owners Group PRA peer review in 2001. Facts and Observations (F&Os) concerning the NOA model were recorded and prioritized to be addressed in the following model update or to be considered as a future model enhancement. In December, 2005 the N05A model was released to address F&Os from the peer review and provide additional model enhancements. A self assessment was performed on this model in August, 2007 to assess the N05A model against the American Society of Mechanical Engineering (ASME) PRA standard RA-Sb-2005
Serial No.10-050 Pg 4 of 26 and Revision 1 of the NRC Regulatory Guide 1.200. This self assessment identified the PRA modeling and documentation supporting requirements (SRs) where the N05A model did not meet Category \I of the ASME PRA Standard.
In December, 2009, the North Anna PRA model N009A was released. This model update was performed to meet the Category \I supporting requirements of the ASME PRA standard and Regulatory Guide 1.200, Rev. 1. The model was reviewed and found to meet all of the recommended modeling supporting requirements appropriate for applying an ASME Code Case N-716 RI-ISI program as identified in the Electric Power Research Institute (EPRI) Topical Report 1018427. Based on the above, Dominion believes that the current North Anna PRA model, which was used to support the Code Case N-716 RI-ISI program, has an acceptable level of quality to support this application.
- 2. PROPOSED ALTERNATIVE TO CURRENT lSI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components.
The alternative RIS_B Program for piping is described in Code Case N-716. The RIS_B Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.
2.2 Augmented Programs The impact of the RIS_B application on the various plant augmented inspection programs listed below were considered. This section documents only those plant augmented inspection programs that address common piping with the RIS_B application scope ( Le., Class 1 and 2 piping).
- The plant augmented inspection program for high-energy line breaks, implemented in accordance with NAPS Updated Final Safety Analysis Report (UFSAR) Section 3C.2.7 "Effect of Piping System Breaks Outside Containment," Table 3C-4 "Break Locations - Main Steam" and Table 3C-7 "Break Locations - Feedwater." This augmented inspection program is governed by procedure ER-NA-AUG-1 01, "North Anna Augmented Inspection Program" and is unaffected by this RIS_B Program.
Selections required by the Code Case N-716 application for the break exclusion region criteria were made to overlap already scheduled exams for the Augmented Program. Examinations will be performed to meet the program specific requirements for which it is being credited.
- The plant augmented inspection program for flow accelerated corrosion (FAG) per Generic Letter (GL) 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, which is relied upon to manage this damage mechanism, is not affected or changed by the RIS_B Program.
Serial No.10-050 Pg 5 of 26
- 3. RISK-INFORMED I SAFETY-BASED lSI PROCESS The process used to develop the RIS_S Program conformed to the methodology described in Code Case N-716 and consisted of the following steps:
- Safety Significance Determination
- Failure Potential Assessment
- Element and NDE Selection
- Risk Impact Assessment
- Implementation Program
- Feedback Loop 3.1 Safety Significance Determination The systems assessed in the RIS_S Program are provided in Tables 3.1a and 3.1b. The piping and instrumentation diagrams and additional plant information including the existing plant lSI Program were used to define the piping system boundaries.
Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are used to determine the treatment requirements. High Safety Significant (HSS) welds are determined in accordance with the requirements below. Low Safety Significant (LSS) welds include all other Class 2, 3, or Non-Class welds.
(1) Class 1 portions of the reactor coolant pressure boundary (RCPS), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii);
(2) Applicable portions of the shutdown cooling pressure boundary function. That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:
(a) As part of the RCPS from the reactor pressure vessel (RPV) to the second isolation valve (Le., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (Le.,
farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; (3) That portion of the Class 2 feedwater system [> 4 inch nominal pipe size (NPS)] of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve; (4) Piping (> NPS 4) within the break exclusion region (SER) for high-energy piping systems as defined by the Owner. This may include Class 3 or Non-Class piping; and (5) Any piping segment whose contribution to CDF is greater than 1E-06 or 1E-07 for LERF based upon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping.
3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant-specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in
Serial NO.1 0-050 Pg 6 of 26 EPRI TR-112657 (i.e., the EPRI RI-ISI methodology), with the exception of the deviation discussed below.
Tables 3.2a and 3.2b summarizes the failure potential assessmentby system for each degradation mechanism that was identified as potentially operative.
A deviation to the EPRI RI-ISI methodology has been implemented in the failure potential assessment for NAPS. Table 3-16 of EPRI TR-112657 contains criteria for assessing the potential for thermal stratification, cycling, and striping (TASeS). Key attributes for horizontal or slightly sloped piping greater than NPS 1 include:
- 1. The potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids; or
- 2. The potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids; or
- 3. The potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid; or
- 4. The potential exists for two phase (steam/water) flow; or
- 5. The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow AND AND Richardson Number> 4 (this value predicts the potential buoyancy of a stratified flow)
These criteria, based on meeting a high cycle fatigue endurance limit with the actual ~T assumed equal to the greatest potential ~T for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TAses where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology that would allow consideration of fatigue severity is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing TASeS susceptibility criteria is presented below.
- Turbulent Penetration TASCS Turbulent penetration typically occurs in lines connected to piping containing hot flowing fluid.
In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic
~Ts can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASeS is considered for this configuration.
For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will keep the line filled with hot water. If there is no potential for in-leakage towards the hot fluid source from the outboard end of the line, this will result in a well-mixed fluid
Serial No.10-050 Pg 7 of 26 condition where significant top-to-bottom ~Ts will not occur. Therefore, TASCS is not considered for these configurations. Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of TASCS will not be significant under these conditions and can be neglected.
- Low flow TASCS In some situations, the transient startup of a system (e.g., shutdown cooling suction piping) creates the potential for fluid stratification as flow is established. In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (TT) will govern.
- Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a significant temperature difference. However, since this is generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected.
- Convection Heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection. However, since there is no potential for cyclic temperature changes in this case, the effect of TASCS is not significant and can be neglected.
In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of TASCS provide an allowance for considering cycle severity. The above criteria have previously been submitted by EPRI to the NRC for generic approval [letters dated February 28,2001, and March 28, 2001, from P.J. O'Regan (EPRI) to Dr. B. Sheron (USNRC), Extension of Risk-Informed Inservice Inspection Methodology]. The methodology used in the RBS RIS_B application for assessing TASCS potential conforms to these updated criteria. Final Materials Reliability Program (MRP) guidance on the subject of TASCS will be incorporated into the RBS RIS_B application, if warranted. It should be noted that the NRC has granted approval for RI-ISI relief requests incorporating these TASeS criteria at several facilities, including Comanche Peak (NRC letter dated September 28,2001) and South Texas Project (NRC letter dated March 5,2002).
Serial No.1 0-050 Pg 8 of 26 3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RIS_B pilot applications provide the criteria for identifying the number and location of required examinations. Ten percent of the HSS welds shall be selected for examination as follows:
(1) Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements:
(a) A minimum of 25% of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.
(b) If the examinations selected above exceed 10% of the total number of HSS welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least 10% of the HSS population is inspected.
(c) If the examinations selected above are not at least 10% of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10%.
(2) At least 10% of the Reactor Coolant Pressure Boundary (RCPB) welds shall be selected.
(3) For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (Le., isolation valve closest to the Reactor Pressure Vessel (RPV)) and the RPV.
(4) A minimum of 10% of the welds in that portion of the RCPB that lies outside containment (OC)
(e.g., portions of the main feedwater system in BWRs) shall be selected..
(5) A minimum of 10% of the welds within the break exclusion region (BER) shall be selected.
In contrast to a number of RI-ISI Program applications where the percentage of Class 1 piping locations selected for examination has fallen substantially below 10%, Code Case N-716 mandates that 10% be chosen. A brief summary is provided below, and the results of the selections are presented in Tables 3.3a and b. Section 4 of EPRI TR-112657 was used as guidance in determining the examination requirements for these locations.
Class 2 Welds(2)
Selected Selected 1 1406 159 1463 2 73 17 2939 178 2 2159 162 1536 o 77 21 3772 183 Notes (1) Includes all Category B-F and B-J locations.
(2) Includes all Category C-F-1 and C-F-2 locations.
(3) NC - non-class (4) Regardless of safety significance, Class 1, 2 and 3 in-scope piping components will continue to be pressure tested as required by the ASME Section XI Program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RIS_B Program.
Serial No.10-050 Pg 9 of 26 3.3.1 Additional Examinations If the flaw is original construction or otherwise is acceptable, Code rules do not require any additional inspections. Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3500 and/or IWB-3600. As part of performing an evaluation to IWB-3600, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation. The process for ordinary flaws is to perform the evaluation using ASME Section XI. If the flaw meets the criteria, then it is noted and appropriate successive examinations scheduled. If the nature and type of the flaw is service-induced, then similar systems or trains will be examined. If the flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000 and/or applicable ASME Section XI Code Cases. The need for extensive root cause analysis beyond that required for IWB-3600 evaluation is dependent on practical considerations (Le., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage). The NRC is involved in the process at several points. For preemptive weld overlays, a relief request in accordance with 10 CFR 50.55a is usually required for design and installation. Should a flaw be discovered during an examination, a notification in accordance with 10 CFR 50.72 or 10 CFR 50.73 may be required.IWB-3600 requires the evaluation to be submitted to the NRC.
Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI.
The evaluation will include whether other elements in the segment or additional segments are subject to the same root cause conditions. Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include HSS elements up to a number equivalent to the number of elements required to be inspected during the current outage. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined during the current outage. No additional examinations need be performed if there are no additional elements identified as being susceptible to the same root cause conditions.
3.3.2 Program Relief Requests An attempt has been made to select RIS_B locations for examination such that a minimum of
>90% coverage (Le., Code Case N-460 criteria) is attainable. However, some limitations will not be known until the examination is performed since some locations may be examined for the first time by the specified techniques. In instances where locations at the time of the examination fail to meet the >90% coverage requirement, the relief process outlined in 10 CFR 50.55a will be followed.
Per footnote 3 of Table 1 of Code Case N-716, when the required examination volume or area cannot be examined due to interference by another component or part geometry, limited examinations shall be evaluated for acceptability. Acceptance of limited examinations or volumes shall not invalidate the results of the change-in-risk evaluation (paragraph 5 of Code Case N-716). The change in risk evaluation of Code Case N-716 is consistent with previous RI lSI applications and meets RG 1.174 change-in-risk acceptance criteria. Areas with acceptable limited examinations, and their bases, shall be documented.
Consistent with previously approved RI-ISI submittals, NAPS will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section XI examinations. Experience has shown this process to be weld-specific (e.g., joint configuration).
As such, the effect on risk, if any, will not be known until that time. Relief requests will be
Serial No.10-050 Pg 10 of 26 submitted per the requirements of 10 CFR 50.55a(g)(5)(iv) within one (1) year after the end of the interval.
3.4 Risk Impact Assessment The RIS_B Program development was conducted in accordance with RG 1.174 and the requirements of Code Case N-716, and the risk associated with implementation of this program is expected to decrease when compared to that estimated from ASME Section XI requirements.
This evaluation categorized segments as HSS or LSS in accordance with Code Case N-716, and then determined what inspection changes were proposed for each system. The changes include changing the number and location of inspections and in many cases improving the effectiveness of the inspection to account for the findings of the RIS_B degradation mechanism assessment. For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.
3.4.1 Quantitative Analysis NAPS Units 1 and 2 have conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716. The analysis estimates the net change in risk due to the positive and negative influences of adding and removing locations from the inspection program. This analysis was performed to ensure that the change in risk of implementing the RIS_B Program meets the requirements of RG 1.174 and 1.178. The Code Case criterion requires that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, respectively. The Code Case also requires that the cumulative increase in overall CDF and LERF be less than 1E-6 and 1E-7 per year, respectively, for the implementation of the RIS-B program.
In accordance with the Code Case, bounding estimates for pipe failure frequency, conditional core damage probability (CCDP), and conditional large early release probability (CLERP) were used to simplify the risk impact assessment calculations. Welds susceptible to FAC were assumed to be managed by the plant FAC augmented inspection program. All welds susceptible to an identified degradation mechanism besides FAC were assigned the failure frequency of 2E-7/yr. All welds that had no identified degradation mechanisms were assigned a failure frequency 1E-8/yr. This approach is consistent with Table 3 of the Code Case.
With respect to assigning failure potential for LSS piping, the criteria are defined by Table 3 of the Code Case. That is, those locations identified as susceptible to FAC (or another mechanism and also susceptible to water hammer) are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion or stress corrosion cracking are assigned to a medium failure potential, and those locations that are identified as not susceptible to degradation are assigned a low failure potential.
Bounding CCDP and CLERP values were conservatively used for all welds in the RIS_B program for the risk impact assessment. Use of these bounding values simplifies the risk impact assessment, and demonstrates that the use of the RIS_B program does not have an adverse affect on safety. Applying bounding CCDPs and CLERPs to all welds is conservative in all cases where the RIS_B program does not increase the number of inspections over the Section XI lSI program. For North Anna Units 1 and 2, the number of inspections increases in the Charging (CH) and the Main Feedwater (FW) Systems.
It is understood that the risk associated with these systems would decrease and would therefore meet the delta risk gUidelines for systems as described in the Code Case. However, to avoid non-conservatism by crediting too much risk reduction from the additional inspections, the total risk impact of the RIS_B program is considered without any risk reduction credited from the CH and FW systems.
Serial No.10-050 Pg 11 of 26 The CCDP and CLERP values used to assess risk impact were estimated based on the potential consequences of pipe break locations under consideration. For CDF, the most limiting case considered was a large break loss of cooling accident. The CCDP for a large LOCA was estimated to be 1.7E-3 for NAPS1 and 2. For LERF, the most limiting case considered was an interfacing systems LOCA (ISLOCA). The CCDP and CLERP for an ISLOCA were estimated by multiplying the CCDP and CLERP estimated by the PRA model by the probability of a check valve to fail open and allow the ISLOCA to take place. The resulting CLERP for an ISLOCA was estimated to be 4.7E-5. These bounding values were conservatively applied to all welds in the RIS_B program for the risk impact assessment.
The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as Xo and is expected to have a value less than 1E-08. Piping locations identified as medium failure potential have a likelihood of 20xo . These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-112657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RIS_B approach.
Tables 3.4a and b present a summary of the RIS_B Program versus the ASME Section XI 1983 Edition through summer 83 Addenda for NAPS1 and the 1986 Edition for NAPS2 program requirements on a "per system" basis. The presence of FAC was adjusted for in the quantitative analysis by excluding their impact on the failure potential rank. The exclusion of the impact of FAC on the failure potential rank and therefore in the determination of the change in risk is appropriate, because FAC is a damage mechanism managed by separate, independent plant augmented inspection programs. The RIS_B Program credits and relies upon these plant augmented inspection programs to manage these damage mechanisms. The plant FAC Program will continue to determine where and when examinations are performed. Hence, since the number of FAC examination locations remains the same "before" and "after" and no delta exists, there is no need to include the impact of FAC in the performance of the risk impact analysis.
As described above, the total risk impact of implementing the RIS_B Program is considered without any credit taken from risk reduction resulting from an increased number of inspections in the CH and FW systems, or from an increased POD. The total change in CDF and LERF for Unit 1 is 1.36E-1 0 and 3.76E-12, respectively, events per year. The total change in CDF and LERF for Unit 2 is -6.97E-1 0 and -1.93E-11, respectively, events per year. These conservative estimates for the change in risk are still below the limits described in the Code Case, and demonstrate that this methodology for risk impact assessment is conservative and appropriate.
As indicated in the following table, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RIS_B Program, and satisfies the acceptance criteria of RG 1.174 and Code Case N-716.
Serial No.10-050 Pg 12 of 26 NAPS Umt . 1 R'ISk Impact Assessment Summary With POD Credit Without POD Credit System Delta CDF Delta LERF Delta CDF Delta LERF CH - Charqinq -1.58E-09 -4.37E-11 -9.01E-10 -2.49E-11 FW - Main Feedwater -2.55E-11 -7.05E-13 -2.55E-11 -7.05E-13 MS - Main Steam 3.40E-11 9.40E-13 3.40E-11 9.40E-13 QS - Quench Spray 1.45E-10 4.00E-12 1.45E-10 4.00E-12 RC - Reactor Coolant -2.05E-09 -5.66E-11 -3.48E-10 -9.64E-12 RH - Residual Heat Removal 9.35E-11 2.59E-12 9.35E-11 2.59E-12 RS - Recirculation Spray 6.80E-11 1.88E-12 6.80E-11 1.88E-12 SI - Safety Injection 1.45E-10 3.99E-12 1.45E-10 3.99E-12 Total -3.17E-09 -8.77E-11 -7.90E-10 -2.19E-11 NAPS Um"t 2 R'ISk ImpactAssessmentS ummary With POD Credit Without POD Credit System Delta CDF Delta LERF Delta CDF Delta LERF CH - Charging -1.97E-09 -5.45E-11 -1.16E-09 -3.20E-11 FW - Main Feedwater -7.65E-11 -2.12E-12 -7.65E-11 -2.12E-12 MS - Main Steam 1.11E-10 3.06E-12 1.11 E-1 0 3.06E-12 QS - Quench Spray 1.53E-10 4.23E-12 1.53E-10 4.23E-12 RC - Reactor Coolant -4.52E-09 -1.25E-10 -1.26E-09 -3.48E-11 RH - Residual Heat Removal 1.62E-10 4.47E-12 1.62E-10 4.47E-12 RS - Recirculation Spray 8.50E-11 2.35E-12 8.50E-11 2.35E-12 SI - Safety Injection 5.10E-11 1.41E-12 5.10E-11 1.41E-12 Total -6.01 E-09 -1.66E-10 -1.93E-09 -5.33E-11 3.4.2 Defense-in-Depth The intent of the inspections mandated by ASME Section XI for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary. Currently, the process for selecting inspection locations is based upon structural discontinuity and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds, this method has been ineffective in identifying leaks or failures.
EPRI TR-112657 and Code Case N-716 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.
This process has two key independent attributes which are a determination of each location's susceptibility to degradation and, secondly, an independent assessment of the consequence of the piping failure. These two attributes assure defense-in-depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, a generic assessment of high-consequence sites has been determined by Code Case N-716 supplemented by plant-specific evaluations, thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or SER break. Finally, Code Case N-716 requires that
Serial No.10-050 Pg 13 of 26 any plant-specific piping with a contribution to CDF of greater than 1E-06 (or 1E-07 for LERF) be included in the scope of the application. No such piping was identified at NAPS.
All locations within the Class 1, 2, and 3 pressure boundaries will continue to be pressure tested in accordance with the Code, regardless of its safety significance.
- 4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RIS_B Program, procedures that comply with the guidelines described in EPRI TR-112657 and Code Case N-716, will be prepared to implement and monitor the program as needed.
The new program will be implemented in the fourth lSI interval. No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.
The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.
The monitoring and corrective action program will contain the following elements:
A. Identify B. Characterize C. (1) Evaluate, determine the cause and extent of the condition identified (2) Evaluate, develop a corrective action plan or plans D. Decide E. Implement F. Monitor G. Trend The RIS_B Program is a living program requiring feedback of new relevant information to ensure the appropriate identification of HSS piping locations. As a minimum, this review will be conducted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or GL requirements, or by industry and plant-specific feedback.
For preservice examinations, NAPS will follow the rules contained in Section 3.0 of Code Case N-716.
Welds classified HSS require preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1 of the Code Case. Welds classified as LSS do not require preservice inspection.
- 5. PROPOSED lSI PROGRAM PLAN CHANGE A comparison between the RIS_B Program and the ASME Section XI 1983 Edition through summer 83 Addenda for NAPS1 and the 1986 Edition for NAPS2 program requirements for in-scope piping is provided in Tables 3.5a and b.
NAPS1 and 2 intend to start implementing the RIS_B Program during the first period of the fourth inspection interval. The fourth lSI interval will implement 100% of the inspection locations selected for examinations per the RIS_B Program. Examinations shall be performed such that the period percentage requirements of ASME Section XI are met.
Serial NO.1 0-050 Pg 14 of 26
- 6. REFERENCES/DOCUMENTATION EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Rev. B-A EPRI TR-1 018427, Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1 Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-Implement Risk-Informed lSI based on ASME Code Case N-716, dated September 21,2007 USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based lSI program for Class 1 and 2 Piping Welds, dated September 28,2007 Supporting Onsite Documentation ET-ISI-2010-0001 Risk Informed Inservice Inspection Programs for NAPS1 and 24th Intervals, Code Case N-716 Based SM-1605 Rev. 0," North Anna 1 and 2 Code Case N-716 Internal Flooding Analysis and Risk Impact Assessment", January 2010 ER-AA-ISI-RI-100 Dominion Risk Informed Program ER-AA-RI-ISI-101 The Dominion Risk Informed Period Update Process ER-AA-ISI-101 Dominion Inservice Inspection Program Preparation and Change Control Process
Serial No.1 0-050 Pg 15 of 26 Table 3.1a N-716 Safety Significance Determination: NAPS1 System Description Weld N-716 Safety Significance Determination Safety Count Significance RCPS SDC PWR: SER >1 E_6 cDF High Low FW >1 E_7 LERF CH - Charging 319 ./ ./
492 ./
81 ./ ./
MS - Main Steam 30 ./ ./
169 ./
RC - Reactor Coolant 570 ./ ./
RH - Residual Heat 31 ./ ./ ./
137 ./
SI - Safety Injection 368 ./ ./
492 ./
QS - Quench Spray 143 ./
RS- Recirculation Spray 73 ./
SUMMARY
RESULTS FOR 1257 ./ ./
ALL SYSTEMS 64 ./ ./
81 ./ ./
31 ./ ./ ./
1506 ./
TOTALS 2939 1433 1506
Serial No.10-050 Pg 16 of 26 Table 3.1b N-716 Safety Significance Determination: NAPS2 System Description Weld N-716 Safety Significance Determination Safety Count Significance RCPB SOC PWR: BER >1 E_6 cDF High Low FW >1 E_7 LERF CH - Charging 420 ./ ./
920 ./
69 ./ ./
21 ./
MS - Main Steam 33 ./ ./
169 ./
RC - Reactor Coolant 573 ./ ./
RH - Residual Heat 29 ./ ./ ./
139 ./
- Safety Injection 360 ./ ./
734 ./
QS - Quench Spray 165 ./
RS- Recirculation Spray 86 ./
SUMMARY
RESULTS FOR 1353 ./ ./
ALL SYSTEMS 77 ./ ./
69 ./ ./
29 ./ ./
2234 ./
TOTALS 3762 1528 2234
Serial No.10-050 Po 17 of 26 Table 3.2a Failure Potential Assessment Summary: NAPS1 System(1) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC 2 ./
2 ./
MS RC ./ ./ ./
2 RH 2 ./ ./
SI 2
QS 2
RS Table 3.2b Failure Potential Assessment Summary: NAPS2 System(1) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC 2 ./
CH 2 ./
FW 2 ./
MS RC ./ ./ ./
2 RH S\2 ./ ./
2 QS 2
RS Notes (1) Systems are described in Tables 3.1 a and 3.1 b (2) A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the QS and RS systems in entirety, as well as portions of the CH, MS, RH, SI systems and only on Unit 2 FW.
Serial No.10-050 Pg 18 of 26 Table 3.3a N-716 Element Selections: NAPS1 System(l) Selections HSS(2) DMs(3) RCPB(4) RCPB 1F1V(5) RCPBO C(6) BER(7)
CH Required 32 of 319 TT 5 of 8 32 of 319 37 n/a n/a Made 51 TT 5 51 49 n/a n/a FIN Required 12 of 115 n/a n/a n/a n/a 5 of 43 Made 13 n/a n/a n/a n/a 10 MS Required 3 of 30 n/a n/a n/a n/a 3 of 30 Made 6 n/a n/a n/a n/a 6 RC Required 57 of 570 TASCS, TT 4 57 of 570 41 n/a n/a TASCS 11 of 41 TT 3 of 11 PWSCC 2 of7 PWSCC, TASCS 1 of 2 Made 62 TASCS, TT 7 62 62 n/a n/a TASCS 5 TT 4 PWSCC 4 PWSCC, TASCS 2 RH Required 3 of 31 n/a 3 of 31 3 n/a n/a Made 4 n/a 4 2 n/a n/a SI Required 37 of 368 IGSCC 2 of 6 37 of 368 0 n/a n/a TT, IGSCC 1 of 3 Made 42 IGSCC 2 42 0 n/a n/a TT,IGSCC 1 TOTAL Made 178 30 159 113 n/a 16 Notes (1) Systems are described in Tables 3.1 a and 3.1 b.
(2) High Safety Significant (3) Degradation Mechanisms No more than 10% of HSS piping welds are required to be selected for examination. DM selections may be reduced to meet this requirement.
(4) Reactor Coolant Pressure Boundary 1F1V 1F1V (5) For RCPB (Reactor Coolant Pressure Boundary inside first isolation valve) 2/3 requirement is for total of RPCB and is not required to be met per system.
(6) Reactor Coolant Pressure Boundary outside containment (7) Break Exclusion Region
Serial No.1 0-050 Pg 19 of 26 Table 3.3b N-716 Element Selections: NAPS2 System(l) Selections HSS(2) DMs(3) RCPB(4) RCPB 1F1V(5) RCPB OC(6) BER(7)
CH Required 42 of 420 TT 5 of 20 42 of 420 33 n/a n/a Made 50 6 50 47 n/a n/a FW Required 11 of 113 n/a n/a n/a n/a 4 of 44 Made 15 n/a n/a n/a n/a 15 MS Required 3 of 33 n/a n/a n/a n/a 3 of 26 Made 6 n/a n/a n/a n/a 6 RC Required 57 of 573 TASCS 12 of 49 57 of 573 45 n/a n/a TASCS, D 4 of 15 D 20f8 PWSCC 1 of 4 D, PWSCC 1 of 2 Made 68 TASCS 17 68 68 n/a n/a TASCS, D 4 D3 PWSCC 1 D, PWSCC 2 RH Required 3 of 29 n/a 3 of 29 3 n/a n/a Made 4 n/a 4 2 n/a n/a SI Required 36 of 360 IGSCC 2 of 6 36 of 360 0 n/a n/a D, IGSCC 1 of 3 Made 40 IGSCC 2 40 0 n/a n/a D,IGSCC 1 TOTAL Made 183 36 162 117 n/a 21 Notes (1) Systems are described in Tables 3.1 a and 3.1 b.
(2) High Safety Significant (3) Degradation Mechanisms No more than 10% of HSS piping welds are required to be selected for examination. OM selections may be reduced to meet this requirement.
(4) Reactor Coolant Pressure Boundary (5) For RCPB 1F1V (Reactor Coolant Pressure Boundary inside first isolation valve) 2/3 requirement is for total of RPCB 1F1V and is not required to be met per system.
(6) Reactor Coolant Pressure Boundary outside containment (7) Break Exclusion Region
10-050 Pg 20 of 26 Table 3.4a Risk Impact Analysis Results: NAPS1 Safety Failure Potential Inspections CDF Impact LERF Impact System(1)
Significance DMs Rank(2) SXI(3) RIS_B Delta wI POD wlo POD wI POD wlo POD CH Hiqh TT Medium 0 5 5 -1.53E-09 -8.50E-10 -4.23E-11 -2.35E-11 CH Hiqh None Low 1 46 45 -3.83E-10 -3.83E-10 -1.06E-11 -1.06E-11 CH Low None Low 39 0 -39 3.32E-10 3.32E-10 9.17E-12 9.17E-12 TOTAL -1.58E-09 -9.01 E-10 -4.37E-11 -2.49E-11 FW Hiqh None Low 10 13 3 -2.55E-11 -2.55E-11 -7.05E-13 -7.05E-13 TOTAL -2.55E-11 -2.55E-11 -7.05E-13 -7.05E-13 MS High None Low 0 6 6 -5.10E-11 -5.10E-11 -1.41E-12 -1.41 E-12 MS Low None Low 10 0 -10 8.50E-11 8.50E-11 2.35E-12 2.35E-12 TOTAL 3.40E-11 3.40E-11 9.40E-13 9.40E-13 QS Low None Low 17 0 -17 1.45E-10 1.45E-10 4.00E-12 4.00E-12 TOTAL 1.45E-10 1.45E-10 4.00E-12 4.00E-12 RC Hiqh PWSCC Medium 5 4 -1 1.70E-10 1.70E-10 4.70E-12 4.70E-12 RC Hiqh PWSCC, TT Medium 2 2 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RC High TASCS Medium 3 5 2 -1.22E-09 -3.40E-10 -3.38E-11 -9.40E-12 RC HiQh TASCS, TT Medium 5 7 2 -3.40E-10 -3.40E-10 -9.40E-12 -9.40E-12 RC Hiqh TT Medium 4 4 0 -8.16E-10 O.OOE+OO -2.26E-11 O.OOE+OO RC HiQh None Low 59 40 -19 1.62E-10 1.62E-10 4.47E-12 4.47E-12 TOTAL -2.05E-09 -3.49E-10 -5.66E-11 -9.64E-12 RH HiQh None Low 1 4 3 -2.55E-11 -2.55E-11 -7.05E-13 -7.05E-13 RH Low None Low 14 0 -14 1.19E-10 1.19E-10 3.29E-12 3.29E-12 TOTAL 9.35E-11 9.35E-11 2.59E-12 2.59E-12 RS Low None Low 8 0 -8 6.80E-11 6.80E-11 1.88E-12 1.88E-12 TOTAL 6.80E-11 6.80E-11 1.88E-12 1.88E-12 SI Hiqh IGSCC Medium 0 2 2 -3.40E-10 -3.40E-10 -9.40E-12 -9.40E-12 SI HiQh IGSCC, TT Medium 0 1 1 -1.70E-10 -1.70E-10 -4.70E-12 -4.70E-12 SI High None Low 61 39 -22 1.87E-10 1.87E-10 5.17E-12 5.17E-12 SI Low None Low 55 0 -55 4.68E-10 4.68E-10 1.29E-11 1.29E-11 TOTAL 1.45E-10 1.45E-10 4.00E-12 4.00E-12 GRAND TOTAL -3.17E-09 -7.91E-10 -8.77E-11 -2.19E-11
Serial No.10-050 Pg 21 of 26 Table 3.4b Risk Impact Analysis Results: NAPS2 Safety Failure Potential Inspections CDF Impact LERF Impact System(1)
Significance DMs Rank(2) SXI(3) RIS_B Delta wI POD w/o POD wI POD w/o POD CH Hioh TT Medium 0 6 6 -1.84E-09 -1.02E-09 -5.08E-11 -2.82E-11 CH Hiqh None Low 1 44 43 -3.66E-10 -3.66E-10 -1.01 E-11 -1.01 E-11 CH Low None Low 27 0 -27 2.30E-10 2.30E-10 6.35E-12 6.35E-12 TOTAL -1.97E-09 -1.16E-09 -5.45E-11 -3.20E-11 FW Hiqh None Low 6 15 9 -7.65E-11 -7.65E-11 -2.12E-12 -2.12E-12 FW Hiqh None Low 0 0 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO TOTAL -7.65E-11 -7.65E-11 -2.12E-12 -2.12E-12 MS Hiqh None Low 0 6 6 -5.10E-11 -5.10E-11 -1.41 E-12 -1.41E-12 MS Low None Low 19 0 -19 1.62E-10 1.62E-10 4.47E-12 4.47E-12 TOTAL 1.11 E-1 0 1.11 E-1 0 3.06E-12 3.06E-12 OS Low None Low 18 0 -18 1.53E-10 1.53E-10 4.23E-12 4.23E-12 TOTAL 1.53E-10 1.53E-10 4.23E-12 4.23E-12 RC Hiqh PWSCC Medium 4 1 -3 5.10E-10 5.10E-10 1.41 E-11 1.41E-11 RC Hiqh PWSCC, TT Medium 2 2 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RC Hioh TASCS Medium 4 17 13 -4.79E-09 -2.21 E-09 -1.33E-10 -6.11E-11 RC Hiqh TASCS, TT Medium 4 4 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO RC Hioh TT Medium 4 3 -1 -5.10E-10 1.70E-10 -1.41 E-11 4.70E-12 RC Hiah None Low 73 41 -32 2.72E-10 2.72E-10 7.52E-12 7.52E-12 TOTAL -4.52E-09 -1.26E-09 -1.25E-10 -3.48E-11 RH High None Low 6 4 -2 1.70E-11 1.70E-11 4.70E-13 4.70E-13 RH Low None Low 17 0 -17 1.45E-10 1.45E-10 4.00E-12 4.00E-12 TOTAL 1.62E-10 1.62E-10 4.47E-12 4.47E-12 RS Low None Low 10 0 -10 8.50E-11 8.50E-11 2.35E-12 2.35E-12 TOTAL 8.50E-11 8.50E-11 2.35E-12 2.35E-12 SI Hiqh IGSCC Medium 0 2 2 -3.40E-10 -3.40E-10 -9.40E-12 -9.40E-12 SI Hiah IGSCC, TT Medium 1 1 0 O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO SI Hioh None Low 24 37 13 -1.11 E-1 0 -1.11 E-1 0 -3.06E-12 -3.06E-12 SI Low None Low 59 0 -59 5.02E-10 5.02E-10 1.39E-11 1.39E-11 TOTAL 5.10E-11 5.10E-11 1.41 E-12 1.41E-12 GRAND TOTAL -6.01 E-09 -1.93E-09 -1.66E-10 -5.33E-11
Serial No.10-050 Pg 22 of 26 Notes (1 ) Systems are described in Tables 3.1 a and b.
(2) The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms.
(3) Only those ASME Section XI Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-112657.
Serial No.10-050 Pg 23 of 26 Table 3.5a Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716: NAPS1(4)
System(1) Safety Failure Potential Code Weld Count Section XI Code Case N-716 Significance Category High Low DMs Rank(2) Vol/Sur Sur Only RIS_B Other(3)
CH ./ TT Medium B-J 18 0 2 5 -
CH ./ None Low B-J 301 1 76 46 -
CH ./ None Low B-J 492 39 12 0 -
C-F-1 14 FW ./ None Low NC(4) 43 0 0 11 -
FW ./ None Low C-F-2 72 10 0 2 -
MS ./ None Low NC 30 0 0 6 -
MS ./ None Low NC 169 10 0 0 -
RC ./ TT Medium B-J 11 4 2 4 -
RC ./ TASCS, TT Medium B-J 16 5 0 7 -
RC ./ TASCS Medium B-J 41 3 9 5 -
RC ./ PWSCC Medium B-F 7 5 0 4 -
RC ./ PWSCC, TT Medium B-F 2 2 0 2 RC ./ None Low B-J 493 59 78 40 -
RH ./ None Low B-J 31 1 0 4 -
RH ./ None Low C-F-1 137 14 0 0 -
SI ./ IGSCC Medium B-J 6 0 0 1 -
SI ./ IGSCC, TT Medium B-J 3 0 0 2 -
Serial No.10-050 Pg 24 of 26 System(1) Safety Failure Potential Code Weld Count Section XI Code Case N-716 Significance Category High Low OMs Rank\"} Other(3)
Vol/Sur Sur Only RIS_B SI ./ None Low B-J 359 61 42 39 -
Sl ./ None Low B-J 492 55 0 0 -
C-F-1 19 QS ./ None Low C-F-1 143 17 0 0 -
RS ./ None Low C-F-1 73 8 0 0 --
Notes (1 ) Systems are described in Tables 3.1 a and b.
(2) The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms. [Note: LSS locations were conservatively assumed to be a rank of Medium (Le., "Assume Medium").]
(3) The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) to be credited toward the 10% requirement. NAPS did not take credit for the FAC Program in meeting the 10% sampling requirement. The "Other" column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals.
(4) The Break Location was not designated in this application because the risk impact assessment was performed with bounding estimates of weld failure consequences in place of weld specific consequences.
Serial No.10-050 Pg 25 of 26 Table 3.5b Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716: NAPS2(4)
System(1) Safety Failure Potential Code Weld Count Section XI Code Case N-716 Significance Category High Low OMs Rank(2) Vol/Sur Sur Only RIS_B Other(3)
CH ./ TT Medium B-J 20 0 6 6 -
CH ./ None Low B-J 400 1 95 44 -
CH ./ None Low B-J 916 27 13 0 -
C-F-1 4 FW ./ None Low NC(4) 44 0 0 15 -
FW ./ None Low C-F-2 69 6 0 0 -
FW ./ None Low C-F-2 21 0 0 0 -
M8 ./ None Low NC 33 0 0 6 -
M8 ./ None Low C-F-2 169 19 7 0 -
RC ./ TT Medium B-J 8 4 0 3 -
RC ./ TA8C8, TT Medium B-J 15 4 0 4 -
RC ./ TA8C8 Medium B-J 49 4 13 17 -
RC ./ PW8CC Medium B-F 4 4 0 1 -
RC ./ PW8CC, IT Medium B-F 2 1 0 2 -
RC ./ None Low B-J 495 73 81 41 -
RH ./ None Low B-J 29 6 0 4 -
RH ./ None Low C-F-1 139 17 0 0 -
81 ./ IG8CC Medium B-J 6 0 0 2 -
81 ./ IG8CC, TT Medium B-J 3 1 0 1 -
Serial No.10-050 Pg 26 of 26 System(1) Safety Failure Potential Code Weld Count Section XI Code Case N-716 Significance Category High Low OMs Rank(2) Vol/Sur High RIS_B Other(3)
SI ./ None Low B-J 351 24 56 37 -
SI ./ None Low B-J 734 24 0 0 -
C-F-1 11 as ./ None Low C-F-1 165 18 0 0 -
RS ./ None Low C-F-1 86 10 1 0 -
Notes (1 ) Systems are described in Tables 3.1 a and b.
(2) The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms. [Note: LSS locations were conservatively assumed to be a rank of Medium (Le., "Assume Medium").]
(3) The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) to be credited toward the 10% requirement. NAPS did not take credit for the FAC Program in meeting the 10% sampling requirement. The "Other" column has been retained in this table solely for uniformity purposes with other RIS_B application template submittals.
(4) The Break Location was not designated in this application because the risk impact assessment was performed with bounding estimates of weld failure consequences in place of weld specific consequences.