NRC 2002-0059, Submittal of Request for Relief No. 3 Risk-Informed Inservice Inspection Program
| ML021900385 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/03/2002 |
| From: | Webb T Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2002-0059 | |
| Download: ML021900385 (39) | |
Text
Kewaunee Nuclear Power Plant Point Beach Nuclear Plant N490 Highway 42 6610 Nuclear Road Kewaunee, WI 54216-9511 Two Rivers, WI 54241 NMC 920.388.2560 920.755.2321 Committed to Nuclear Excellence Kewaunee / Point Beach Nuclear Operated by Nuclear Management Company, LLC NRC 2002-0059 10 CFR 50.55a July 3, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Dockets 50-266 and 50-301 Point Beach Nuclear Plant, Units 1 and 2 Submittal of Request for Relief No. 3 Risk-Informed Inservice Inspection Program Ladies and Gentlemen:
On July 1, 2002, Point Beach Nuclear Plant (PBNP) updated the Inservice Inspection (ISI)
Program to the 1998 Edition of ASME Section XI with all addenda through 2000. This edition and addenda were approved for use via a safety evaluation report (SER) issued by the Commission dated November 6, 2001. The SER contained a provision that PBNP implement the requirements in the proposed rule for 10 CFR 50.55a dated August 3, 2001. When the final rule is issued, PBNP will be required to update the ISI Program to follow those rules. The date of the fourth interval start date (July 1, 2002) was approved via an NRC SER dated June 18, 2001.
In accordance with 10 CFR 50.55a(a)(3)(i), PBNP requests NRC authorization to use Risk Informed criterion for piping weld selection. This will utilize the guidelines of ASME Code Case N-578 and EPRI Topical Report TR-1 12657. This Risk-Informed Inservice Inspection criterion will provide an acceptable level of quality and safety.
Enclosed is a request for relief to allow performing the Risk-Informed Inservice Inspection Program as an alternative to the requirements of ASME Section XI categories B-F, B-J, C-F-i, and C-F-2 examination methods and selection criterion. The template for the Risk-Informed Inservice Inspection Program is similar to those submitted by the majority of the plants requesting this alternative to the ASME Section XI requirements.
NRC 2002-0059 Page 2 Approval of this relief request is desired by September 14, 2002. Please contact us if there are any questions regarding this relief request.
Sincerely, ThomasJ. Webb Regulatory Affairs Manager LAS/kmd Attachment Enclosure cc:
NRC Regional Administrator NRC Resident Inspector Mike Verhagen NRC Project Manager PSCW
NRC 2002-0059 ISI Relief Request No. 3 Page 1 Relief Request No. 3, Risk Informed Examination of Class 1 and Class 2 Piping Welds (Code Case N-578 and EPRI TR-1 12657)
Pursuant to 10 CFR 50.55a(a)(3)(i), PBNP requests an alternative to the requirements of Code Categories B-F, B-J, C-F-i, and C-F-2 piping welds as specified in the 1998 Edition of ASME Section XI with Addenda through 2000 as modified by NRC SER dated November 6, 2001.
COMPONENT IDENTIFICATION Plant/Unit:
Code Classes:
References:
Examination Categories:
Item Numbers:
==
Description:==
Component Numbers:
CODE REQUIREMENT Point Beach Nuclear Plant, Units 1 and 2 1 and 2 1998 Edition of Section XI with Addenda through 2000 Tables IWB-2500-1 and IWC-2500-1 ASME Section XI Code Case N-578 B-F, B-J, C-F-i, C-F-2 B5.10. B5.20, B5.40, B5.50, B5.70, B9.11, B9.21, B9.31, B9.32, B9.40, C5.11, C5.21, C5.30, C5.41, C5.51, C5.61, C5.70, C5.81 All pressure retaining piping welds See attached Risked-Informed Inservice Inspection Plan, Point Beach Nuclear Plant, Units 1 and 2, Revision 0 The 1998 Edition of Section XI with Addenda through 2000, IWB-2500 (a) states, "components shall be examined and tested as specified in Table IWB-2500-1. The method of examination for the components and parts of the pressure retaining boundaries shall comply with those tabulated in Table IWB-2500-1 except where alternate examination methods are used that meet the requirements of IWA-2240."
Table IWB-2500-1, Categories B-F and B-J requires 100% and 25% respectively of the total number of non-exempt welds.
The 1998 Edition of Section XI with Addenda through 2000, IWC-2500 (a) states, "Components shall be examined and pressure tested as specified in Table IWC-2500-1.
The method of examination for the components and parts of the pressure retaining boundaries shall comply with those tabulated in Table IWC-2500-1, except where alternate examination methods are used that meet the requirements of IWA-2240."
In addition, Tables IWB-2500-1 and IWC-2500-1 reference figures that convey the examination volume for each configuration that could be encountered.
NRC 2002-0059 ISI Relief Request No. 3 Page 2 BASIS FOR ALTERNATIVE The scope for the ASME Section XI ISI Programs is largely based on deterministic results contained in design stress reports. These reports are normally very conservative and may not be an accurate representation of failure potential. Since the stress reports for PBNP do not contain all the information required to select welds in accordance with the later editions of ASME Section XI, PBNP has been utilizing the alternative selection methodology of the 1974 Edition of Section XI with addenda through summer 1975.
Industry service experience has shown that piping weld failures are due to either corrosion or fatigue and typically occur in areas not included in the plant's ISI program.
Consequently, nuclear plants are devoting significant resources to inspection programs that provide minimum benefit.
As an alternative, significant industry attention has been devoted to the application of risk-informed selection criteria in order to determine a more appropriate scope for ISI Programs at nuclear power plants. EPRI studies indicate that the application of Risk Informed techniques will allow operating nuclear plants to reduce the examination scope of current ISI Programs by as much as 60% to 80%, significantly reduce costs and radiation exposure, and continue to maintain an equivalent or better safety level as current ASME Section XI selection criteria.
PBNP has reviewed the EPRI Methodology as documented in the NRC approved EPRI Topical Report TR-1 12657B-A and referenced in Code Case N-578. Utilizing this methodology for the selection and subsequent examination of PBNP Class I and 2 piping welds will provide an acceptable level of quality and safety.
ALTERNATE EXAMINATION As an alternative to existing ASME Section XI requirements for piping weld selection and examination volumes, PBNP will implement the alternative methods as specified in Code Case N-578 and EPRI TR-1 12657B-A. "Risk-Informed Inservice Inspection Program Plan, Point Beach Nuclear Plant, Unit 1 and 2, Revision 0", details the methodology for application of this alternative.
Pressure testing will be performed as required by ASME Section XI, 1998 Edition with Addenda through 2000.
APPLICABLE TIME PERIOD The alternative is requested for the remainder of the fourth 1 0-year Inservice Inspection interval for PBNP, which began July 1, 2002 and ends on June 30, 2013.
ATTACHMENTS Risk-Informed Inservice Inspection Program Plan, Point Beach Nuclear Plant, Unit 1 and 2, Revision 0
NRC 2002-0059 ISI Relief Request No. 3 Enclosure
RISK-INFORMED INSERVICE INSPECTION PROGRAM PLAN POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REVISION 0 Table of Contents
- 1.
Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 PSA Quality
- 2.
Proposed Alternative to Current ISI Program Requirements 2.1 ASME Section XI 2.2 Augmented Programs
- 3.
Risk-Informed ISI Process 3.1 Scope of Program 3.2 Consequence Evaluation 3.3 Failure Potential Assessment 3.4 Risk Characterization 3.5 Element and NDE Selection 3.5.1 Additional Examinations 3.5.2 Program Relief Requests 3.6 Risk Impact Assessment 3.6.1 Quantitative Analysis 3.6.2 Defense-in-Depth
- 4.
Implementation and Monitoring Program
- 5.
Proposed ISI Program Plan Change
- 6.
References/Documentation
- 1.
INTRODUCTION The Point Beach Nuclear Plant (PBNP) is currently nearing the end of its third inservice inspection (ISI) interval as defined by the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Section XI Code for Inspection, Program B. PBNP plans to implement a risk-informed inservice inspection (RI-ISI) program at the start of the fourth inservice inspection interval, which begins July 1, 2002. The ASME Section Xl Code used during the third interval was the 1986 Edition. Pursuant to 10 CFR 50.55a(a)(3)(i), PBNP requested to use the 1998 Edition of Section XI with addenda through 2000 for the fourth inservice inspection interval. The NRC granted this relief request on November 6, 2001 with the stipulation that the proposed rule of August 3, 2001 be followed. When the final rule is issued, PBNP will be required to update the ISI Program to follow those rules. This will be the applicable ASME Section Xl Code for the fourth interval at PBNP.
The objective of this submittal is to request the use of a risk-informed process for the inservice inspection of Class 1 and 2 piping. The RI-ISI process used in this submittal is described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A "Revised Risk Informed Inservice Inspection Evaluation Procedure." The RI-ISI application was also conducted in a manner consistent with ASME Code Case N-578 "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B."
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 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions On Plant-Specific Changes to the Licensing Basis" and Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping". Further information is provided in Section 3.6.2 relative to defense-in-depth.
1.2 PSA Quality Summary of Probabilistic Risk Assessment (PRA) Level 1 Results: The Point Beach Level 1 PRA 2001 calculated a core damage frequency (CDF) for internal events of 4.4E-5/year for each PBNP unit. Among the contributors to the PBNP CDF, no single initiator clearly dominates. However, the largest contributor, with a 20% contribution, is Steam Generator Tube Rupture (SGTR). The SGTR core damage sequences are largely dominated by human error events.
The Point Beach internal flooding CDF (1.08E-5/year) is dominated by two flooding sources: (1) Rupture of a service water header in the auxiliary building and (2) Rupture of a circulating water expansion joint or fire water main in the water intake facility pump house.
Summary of Level 2 PRA Results: The Point Beach Level 2 PRA calculated a Large Early Release Frequency (LERF) of 1.2E-5/year for each PBNP unit, resulting in a LERF/CDF ratio of 0.27. The calculated LERF is conservative, since all SGTR core damage sequences, and all induced SGTR sequences (from Steam Line and Feedwater breaks), are assumed to contribute to LERF.
Page 2 of 34
PRA Model
Description:
A detailed Level 1 PRA of PBNP was performed in accordance with the methodology described in NUREG/CR-2300, "PRA Procedures Guide." The Point Beach PRA models were developed using small event trees (primarily systemic) and large fault trees. The model represents accident and transient initiating events starting from power operation and continuing for a 24-hour mission time. The PBNP design modeled in the PRA and used in the Individual Plant Examination (IPE) was based on a design freeze date of September 5, 1990. Therefore, the quantification results presented in the IPE report reflect the September 1990 design and operation of PBNP. The IPE was submitted to the NRC on June 30, 1993. This submittal and supporting documentation was reviewed by internal PRA and systems and operations experts and by independent industry PRA specialists. The NRC issued a Staff Evaluation Report (SER) on the Point Beach IPE on January 26, 1995.
Several design modifications have been installed since the IPE PRA model was developed. A 1993 update to the PRA model incorporated installation of an additional safety-related station battery, a non-safety-related battery, and alternate shutdown switchgear with associated 13.8 KV system upgrades, and upgrades of the Gas Turbine Generator and the Main Steam Isolation Valves to improve reliability. The installation of two additional emergency diesel generators and a plant specific data update were incorporated into a 1996 revision of the model. The original results provided for the RI-ISI analysis are based on the 1996 PRA Update. The last PRA update was done in the summer of 2001, to prepare for the PRA Certification effort. To the degree available, those results are incorporated into this study.
The internal flooding analysis for PBNP was conducted by dividing the plant into six major flooding zones and evaluating the sources and effects of flooding in each of the zones. Submergence, spray, dripping, and steam blanketing effects resulting from ruptures/leaks of piping/gaskets, valves, pumps, expansion joints, heat exchangers, and tanks were addressed.
The limited-scope Level 2 PRA of PBNP was performed using the dominant Level 1 core melt sequences binned into 17 plant damage states (including bypass sequences).
Containment response and radioactive source terms for these plant damage states were determined with MAAP 3.0B (PWR Version 19) analysis for a 48-hour mission time.
Certain phenomenological issues were addressed using PBNP-specific position papers.
A best-estimate containment failure pressure fragility curve was calculated for PBNP.
Containment Event Trees (CETs) were used to characterize the containment response to core melt sequences, including uncertainties, using plant damage states quantified with Level 1 PRA models updated to include containment cooling systems. A separate containment isolation fault tree was developed to quantify the probability of containment isolation failure.
Current Point Beach PRA Model: The most recent update of the PBNP PRA model was completed during the summer of 2001. Updates included initiating event frequencies, event trees, failure data, and system models for the four most risk significant systems:
auxiliary feedwater, service water, ECCS (safety injection and residual heat removal),
and electric power. A partial update to human reliability analysis was also performed.
Finally, the event tree / fault tree logic was converted to an all fault tree top logic. This enables a direct upload of the model to the risk monitoring software Safety Monitor. As a result, future model updates can be more easily incorporated into Safety Monitor to Page 3 of 34
keep it current with plant changes. The Point Beach PRA model update process is controlled by ESG 5.1, "PRA Maintenance and Update Guideline."
WOG Peer Review: This current PRA model version was reviewed in June 2001 by a Westinghouse Owners Group PRA Peer Review Team. The team consisted of a team leader from Westinghouse, two contract PRA reviewers, and three reviewers from PRA groups at other Westinghouse power plants. In general, the review team concluded that the PBNP PRA could be effectively used to support applications involving risk significance determinations supported by deterministic analyses once the items noted in the report are addressed. A major observation was that the thermal hydraulic bases for system and human action success were largely from either conservative design basis analyses or analyses that were not specific to Point Beach. These thermal hydraulic bases date from the original PRA done for the IPE. Other observations discussed shortcomings with the basis and documentation of the common cause failure analysis, a general lack of miscalibration errors in the model, the need to complete the human reliability analysis update, and the need to document the quantification and complete documentation of the remainder of the model. Some of these items (primarily documentation issues) have already been addressed since the Peer Review team completed their work. The thermal hydraulic analysis items will take considerable work to resolve, and these are scheduled for 2002. While addressing the Peer Review Teams observations will take time to resolve completely, they are largely concerned with providing a well documented basis for the model, and are not expected to result in model changes that will significantly affect the overall results or conclusions of the Risk Informed ISI consequence evaluation.
- 2.
PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAM REQUIREMENTS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-1 and C-F-2 currently contain the requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components. The alternative RI-ISI program for piping is described in EPRI TR-1 12657.
The RI-ISI 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 Xl Code will be unaffected. EPRI TR-1 12657 provides the requirements for defining the relationship between the RI-ISI program and the remaining unaffected portions of ASME Section Xl.
2.2 Augmented Programs The following augmented inspection programs were considered during the RI-ISI application:
"* The augmented inspection program for Flow Accelerated Corrosion (FAC) per Generic Letter 89-08 is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RI-ISI program.
"* The augmented inspection program for high-energy break exclusion piping is not affected or changed by the RI-ISI program.
Page 4 of 34
- 3.
RISK-INFORMED ISI PROCESS The process used to develop the RI-ISI program conformed to the methodology described in EPRI TR-1 12657 and consisted of the following steps:
Scope Definition Consequence Evaluation Failure Potential Assessment 0
Risk Characterization Element and NDE Selection 0
Risk Impact Assessment 0
Implementation Program 0
Feedback Loop A deviation to the EPRI RI-ISI methodology has been implemented in the failure potential assessment for PBNP. Table 3-16 of EPRI TR-1 12657 contains criteria for assessing the potential for thermal stratification, cycling and striping (TASCS). Key attributes for horizontal or slightly sloped piping greater than 1" nominal pipe size (NPS) include:
- 1.
Potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids, or
- 2.
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.
Potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid, or
- 4.
Potential exists for two phase (steam/water) flow, or
- 5.
Potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow, AND AT > 50'F, 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 AT assumed equal to the greatest potential AT for the transient, will identify all locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TASCS 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 TASCS susceptibility criteria is presented below.
Page 5 of 34
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 ATs can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS 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 condition where significant top-to-bottom ATs 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., RHR 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 a 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 the consideration of cycle severity in assessing the potential for TASCS effects. The above criteria have previously been submitted by EPRI for generic approval (Letter dated February 28, 2001, P.J. O'Regan (EPRI) to Dr. B.
Sheron (USNRC), "Extension of Risk-Informed Inservice Inspection Methodology").
Page 6 of 34
3.1 Scope of Program The systems included in the RI-ISI program are provided in Tables 3.1-1 and 3.1-2 for Units 1 and 2, respectively. The piping and instrumentation diagrams and additional plant information including the existing plant ISI program were used to define the Class 1 and 2 piping system boundaries.
3.2 Consequence Evaluation The consequence(s) of pressure boundary failures were evaluated and ranked based on their impact on core damage and containment performance (i.e., isolation, bypass and large early release). The impact on these measures due to both direct and indirect effects was considered using the guidance provided in EPRI TR-1 12657.
3.3 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 EPRI TR-1 12657, with the exception of the previously stated deviation.
Tables 3.3-1 and 3.3-2 summarize the failure potential assessment by system for each degradation mechanism that was identified as potentially operative in Units 1 and 2, respectively.
3.4 Risk Characterization In the preceding steps, each run of piping within the scope of the program was evaluated to determine its impact on core damage and containment performance (i.e., isolation, bypass and large, early release) as well as its potential for failure. Given the results of these steps, piping segments are then defined as continuous runs of piping potentially susceptible to the same type(s) of degradation and whose failure will result in similar consequence(s). Segments are then ranked based upon their risk significance as defined in EPRI TR-112657.
The results of these calculations are presented in Tables 3.4-1 and 3.4-2 for Units 1 and 2, respectively.
3.5 Element and NDE Selection In general, EPRI TR-1 12657 requires that 25% of the locations in the high-risk region and 10% of the locations in the medium risk region be selected for inspection using appropriate NDE methods tailored to the applicable degradation mechanism. In addition, per Section 3.6.4.2 of EPRI TR-1 12657, if the percentage of Class 1 piping locations selected for examination falls substantially below 10%, then the basis for selection needs to be investigated.
For PBNP Unit 1, the percentage of Class 1 welds selected per the RI-ISI process was 9.3% (70 of 754 welds), which is not a significant departure from 10%.
Page 7 of 34
One additional factor that was considered during the Unit 1 evaluation was the overall percentage of Class 1 selections included both socket and non-socket welds. Therefore, the final percentage of Class 1 selections was 9.3% when both socket and non-socket piping welds were considered. This percentage increases to 20.3% (66 of 325 welds) when considering only those piping welds that are non-socket welded. It should be noted that non-socket welds are subject to volumetric examination, so this percentage does not rely upon welds that are solely subject to a VT-2 visual examination.
For PBNP Unit 2, the percentage of Class 1 welds selected for examination per the RI-ISI process was 10.1% (63 of 621 welds).
As stated in TR-1 12657, the existing FAC augmented inspection program provides the means to effectively manage this mechanism. No additional credit was taken for any FAC augmented inspection program locations beyond those selected by the RI-ISI process to meet the sampling percentage requirements.
A brief summary is provided in the following table, and the results of the selections are presented in Tables 3.5-1 and 3.5-2. Section 4 of EPRI TR-1 12657 was used as guidance in determining the examination requirements for these locations.
Unt Class I Piping Welds(1 class 2 Piping Welds (2)
All Piping Welds (3)
Total Selected Total Selected Total Selected 1
754 70 1068 58 1822 128 2
621 63 1152 69 1773 132 Notes
- 1.
Includes all Category B-F and B-J locations.
- 2.
Includes all Category C-F-1 and C-F-2 locations.
- 3.
All in-scope piping components, regardless of risk classification, will continue to receive Code required pressure testing, as part of the current ASME Section Xl program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RI-ISI program.
3.5.1 Additional Examinations The RI-ISI program in all cases will determine through an engineering evaluation the root cause of any unacceptable flaw or relevant condition found during examination. The evaluation will include the applicable service conditions and degradation mechanisms to establish that the element(s) will still perform their intended safety function during subsequent operation. Elements not meeting this requirement will be repaired or replaced.
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 high risk significant elements and medium risk significant elements, if needed, up to a number equivalent to the number of elements required to be inspected on the segment or segments 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. No additional examinations Page 8 of 34
will be performed if there are no additional elements identified as being susceptible to the same root cause conditions.
3.5.2 Program Relief Requests An attempt has been made to select RI-ISI locations for examination such that a minimum of >90% coverage (i.e., 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 are found at the time of the examination that do not meet the >90% coverage requirement, the process outlined in EPRI TR-1 12657 will be followed.
None of the existing PBNP relief requests are being withdrawn due to the RI-ISI application.
3.6 Risk Impact Assessment The RI-ISI program has been conducted in accordance with Regulatory Guide 1.174 and the requirements of EPRI TR-1 12657, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.
This evaluation identified the allocation of segments into High, Medium, and Low risk regions of the EPRI TR-1 12657 and ASME Code Case N-578 risk ranking matrix, and then determined for each of these risk classes what inspection changes are proposed for each of the locations in each segment. The changes include changing the number and location of inspections within the segment and in many cases improving the effectiveness of the examination to account for the findings of the RI-ISI degradation mechanism assessment. For example, for locations subject to thermal fatigue, examinations will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the examination process.
3.6.1 Quantitative Analysis Limits are imposed by the EPRI methodology to ensure that the change in risk of implementing the RI-ISI program meets the requirements of Regulatory Guides 1.174 and 1.178. The EPRI criterion requires that the cumulative change in core damage frequency (CDF) and large early release frequency (LERF) be less than 1 E-07 and 1 E-08 per year per system, respectively.
Page 9 of 34
Point Beach conducted a risk impact analysis per the requirements of Section 3.7 of EPRI TR-1 12657. The analysis estimates the net change in risk due to the positive and negative influence of adding and removing locations from the inspection program. A risk quantification was performed using the "Simplified Risk Quantification Method" described in Section 3.7 of EPRI TR-1 12657. The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) used for high consequence category segments was based on the highest evaluated CCDP (2.8E-02) and CLERP (1.1E-03), whereas, for medium consequence category segments, bounding estimates of CCDP (1E-04) and CLERP (1E-05) were used. 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 1 E-08. Piping locations identified as medium failure potential have a likelihood of 20x0. These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-1 12657. 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 RI-ISI approach.
Tables 3.6-1 and 3.6-2 present summaries of the RI-ISI program versus 1986 ASME Section Xl Code Edition program requirements and identify on a per system basis each applicable risk category for Units 1 and 2, respectively. The presence of FAC was adjusted for in the performance of the quantitative analysis by excluding its impact on the risk ranking. However, in an effort to be as informative as possible, for those systems where FAC is present, Tables 3.6-1 and 3.6-2 present the information in such a manner as to depict what the resultant risk categorization is both with and without consideration of FAC. This is accomplished by enclosing the FAC damage mechanism, as well as all other resultant corresponding changes (failure potential rank, risk category and risk rank), in parenthesis. Again, this has only been done for information purposes, and has no impact on the assessment itself. The use of this approach to depict the impact of degradation mechanisms managed by augmented inspection programs on the risk categorization is consistent with that used in the delta risk assessment for the Arkansas Nuclear One, Unit 2 (ANO-2) pilot application. An example is provided on the following page.
Page 10 of 34
Risk Consequence Failure Potential Category Rank"'
Rank DMs Rank In this example if FAC is not considered, the failure potential rank is "medium" instead of "high" based on the TASCS and TT damage mechanisms. When a "medium" failure potential rank is combined with a "medium" consequence rank, it results in risk category 5 ("medium" risk) being assigned instead of risk category 3 ("high" risk).
FW 5(3)
Medium (High)
Medium TASCS, TT, (FAC):
Medium (High)
In this example if FAC were considered, the failure potential rank would be "high" instead of "medium". If a "high" failure potential rank were combined with a "medium" consequence "rank, it would result in risk category 3 ("high" risk) being assigned instead of risk category 5 ("medium" risk).
Note
- 1. The risk rank is not included in Tables 3.6-1 or 3.6-2 but it is included in Tables 5-2-1 and 5-2-2.
As indicated in the following tables, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RI-ISI program, and satisfies the acceptance criteria of Regulatory Guide 1.174 and EPRI TR-1 12657.
Unit I Risk Impact Results teARSkCDF ARiskLERF w/te~
1 wPOD J
-3.21 E-08
-2.94E-09
-1.2611-09
-1.16E-10 AC
-4.19E-09
-8.30E-10
-1.64E-10
-3.20E-1 I SI 2.38E-09 2.38E-09 9.35E-1 1 9.35E-1 1 CVC
-5.60E-10
-5.60E-10
-2.20E-1 1
-2.20E-11 MS 8.40E-10 8.40E-10 3.30E-1 1 3.30E-1 1 FW
-1.18E-08 2.80E-09
-4.62E-10 1.10E-10 AFW
-8.54E-09
-2.94E-09
-3.36E-1s
-1.16E-10 Total
-5.39E-08
-1.25E-09
-2.12E-09
-4.85E-10 Note
- 1.
Systems are described in Table 3.1-1.
Page 11 of 34
Unit 2 Risk Impact Results System(1 ARiSkCDF ARiSkLERF w/ POD w/o POD w/ POD wlo POD RC
-2.03E-08 3.22E-09
-7.98E-10 1.27E-10 AC
-5.61E-09
-2.25E-09
-2.21E-10
-8.90E-11 SI
-3.23E-09
-3.23E-09
-1.28E-10
-1.28E-10 CVC
-5.78E-10
-5.70E-10
-2.38E-11
-2.30E-11 MS no change no change no change no change FW
-1.01 E-08 no change
-3.96E-10 no change AFW
-8.82E-09
-3.22E-09
-3.47E-10
-1.27E-10 Total
-4.86E-08
-6.05E-09
-1.91 E-09
-2.40E-10 Note
- 1. Systems are described in Table 3.1-2.
3.6.2 Defense-in-Depth The intent of the examinations 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 ASME Code process for picking 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. As allowed by 10 CFR 50.55a(b)(2)(ii), examinations may alternatively be performed on a random selection of the Class 1 and 2 piping weld population per the 1974 Edition through 1975 Addenda of ASME Section Xl.
This was the process used at PBNP during their third inservice inspection interval. EPRI TR-1 12657 and Code Case N-578 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.
This process has two key independent ingredients, that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients 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, the consequence assessment effort has a single failure criterion. As such, no matter how unlikely a failure scenario is, it is ranked High in the consequence assessment, and at worst Medium in the risk assessment (i.e., Risk Category 4), if as a result of the failure there is no mitigative equipment available to respond to the event. In addition, the consequence assessment takes into account equipment reliability, and less credit is given to less reliable equipment.
All locations within the Class 1 and 2 pressure boundaries will continue to receive a system pressure test and visual VT-2 examination as currently required by the Code regardless of its risk classification.
Page 12 of 34
- 4.
IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RI-ISI program, procedures that comply with the guidelines described in EPRI TR-1 12657 will be prepared to implement and monitor the program. The new program will be integrated into the fourth inservice inspection interval. No changes to the Technical Specifications or 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 Xl program implementing procedures will be retained and modified to address the RI-ISI 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 RI-ISI program is a living program requiring feedback of new relevant information to ensure the appropriate identification of high safety significant piping locations. As a minimum, risk ranking of piping segments will be reviewed and adjusted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback.
- 5.
PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the RI-ISI program and ASME Section X1 Code 1986 Edition program requirements for in-scope piping is provided in Tables 5-1-1 and 5-2-1 for Unit 1 and Tables 5-1 2 and 5-2-2 for Unit 2. Tables 5-1-1 and 5-1-2 provide summary comparisons by risk region.
Tables 5-2-1 and 5-2-2 provide the same comparison information, but in a more detailed manner by risk category, similar to the format used in Table 3.6-1 and 3.6-2.
PBNP is implementing the RI-ISI program at the start of the first period of its fourth inspection interval. As such, 100% of the required RI-ISI program inspections will be completed in the fourth interval. Examinations shall be performed during the interval such that the period examination percentage requirements of ASME Section XI, paragraphs IWB-2412 and IWC 2412 are met.
Page 13 of 34
- 6.
REFERENCES/DOCUMENTATION EPRI TR-1 12657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure", Rev. B-A ASME Code Case N-578, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B,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" Supporting Onsite Documentation Structural Integrity Calculation/File No. NMC-01-310, "Degradation Mechanism Evaluation for the Point Beach Nuclear Plant (PBNP) - Units 1/2", Revision 4 Structural Integrity Calculation/File No. NMC-01-31 1, "Risk-Informed Inservice Inspection Consequence Evaluation of Class 1 & 2 Piping - Point Beach Nuclear Plant Units 1 & 2",
Revision 2 Structural Integrity Calculation/File No. NMC-01-312, "Risk Ranking Summary, Matrix and Report for the Point Beach Nuclear Plant ", Revision 0 Structural Integrity Calculation/File No. NMC-01-313, "Risk Impact Analysis for the Point Beach Nuclear Plant ", Revision 0 Structural Integrity File No. NMC-01-103-1, Record of Conversation No. ROC-004, "Minutes of the Element Selection Meeting for the Risk-Informed IS[ Project at the Point Beach Nuclear Plant", Revision 0, dated November 29 - 30, 2001 Structural Integrity Calculation/File No. NMC-01-315, "Risk-Informed Inservice Inspection Service History Review", Revision 0 Page 14 of 34
Table 3.1-1 Unit I -System Selection and Segment / Element Definition
System Description
Number of Segments Number of Elements RC - Reactor Coolant 55 350 AC - Auxiliary Coolant 67 424 SI - Safety Injection 72 525 CVC - Chemical and Volume Control 14 231 MS - Main Steam 20 80 FW - Feedwater 14 48 AFW - Auxiliary Feedwater 10 164 Totals 252 1822 Page 15 of 34
Table 3.1-2 Unit 2 - System Selection and Segment / Element Definition
System Description
Number of Segments Number of Elements RC - Reactor Coolant 56 309 AC - Auxiliary Coolant 56 388 SI - Safety Injection 80 577 CVC - Chemical and Volume Control 15 171 MS - Main Steam 24 89 FW - Feedwater 10 52 AFW - Auxiliary Feedwater 10 187 Totals 251 1773 Page 16 of 34
Table 3.3-1 Unit 1 - Failure Potential Assessment Summary System~1)
Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RC X
X AC X
X X
Si x
CVC MS FW X
x AFW X
Note
- 1. Systems are described in Table 3.1-1.
Page 17 of 34
Table 3.3-2 Unit 2 - Failure Potential Assessment Summary System~ 1)
Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RC X
X X
AC X
X X
SI x
CVC X
MS AFW X
Note
- 1.
Systems are described in Table 3.1-2.
Page 18 of 34
High Risk Region Medium Risk Region Low Risk Region System(l)
Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With Without With Without With
_Without With Without With WI WithoutWitth I
With Without RC 13 13 42 42 AC 3
3 36 36 2
2 23 23 3
3 SI 1
1 28 28 40 40 3
3 CVC 4
4 10 10 MS 20 20 FW 2(2) 0 2
4 4(3) 0 6
10 AFW 2
2 8
8 Total 2
0 21 23 4
0 138 138 2
2 79 83 6
6 Notes
- 1.
Systems are described in Table 3.1-1.
- 2.
These two segments become Category 2 after FAC is removed from consideration due to the presence of another "medium" failure potential damage mechanism.
- 3.
These four segments become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.
Page 19 of 34 Table 3.4-1 Unit I - Number of Segments by Risk Category With and Without Impact of FAC
Table 3.4-2 Unit 2 - Number of Segments by Risk Category With and Without Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(1 )
Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With Without With Without With Without With Without With Without With Without With Without RC 15 15 41 41 AC 3
3 32 32 1
1 17 17 3
3 SI 1
1 32 32 1
1 43 43 3
3 CVC 4
4 1
1 10 10 MS 24 24 FW 2(2) 0 1
3 2(3) 0 5
7 AFW 2
2 8
8 Total 2
0 22 24 2
0 141 141 3
3 75 77 6
6 Notes
- 1. Systems are described in Table 3.1-2.
- 2. These two segments become Category 2 after FAC is removed from consideration due to the presence of another "medium" failure potential damage mechanism.
- 3. These two segments become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.
Page 20 of 34
Table 3.5-1 Unit I - Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(1 )
Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 Total Selected Total Selected Total Selected Total Selected Total Selected Total Selected Total Selected RC 33 9
317 32 AC 4
1 297 30 4
1 115 0
4 0
SI 2
1 198 20 305 0
20 0
CVC 32 4
199 0
MS 80 8
FW 11 4(2) 37 0
AFW 6
2 158 16 Total 56 17 1082 110 4
1 656 0
24 0
Notes
- 1. Systems are described in Table 3.1-1.
subject to both FAC and RI-ISI examinations.
Since a damage mechanism other than FAC was identified, these welds will be Page 21 of 34
Table 3.5-2 Unit 2 - Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(1 )
Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 Total Selected Total Selected Total FSelected Total Selected TotalTSelected Total Selected Total Selce RC 32 8
277 28 AC 4
1 274 28 2
1 104 0
4 0
SI 2
1 270 27 1
1 279 0
25 0
CVC 36 4
1 1
134 0
MS 89 9
FW 8
3(2) 44 0
AFW 8
2 179 18 Total 54 15 1125 114 4
3 561 0
29 0
Notes
- 1.
Systems are described in Table 3.1-2.
- 2.
Two of the three welds were selected for examination by both the FAC and RI-ISI Programs.
subject to both FAC and RI-ISI examinations.
Since a damage mechanism other than FAC was identified, these welds will be Page 22 of 34
Table 3.6-1 Unit I - Risk Impact Analysis Results System(1 )
Category Consequence Failure Potential Inspections CDF Impact( 3 )
LERF Impacte3)
Rank DMs Rank Section XI(2)
RI-ISI Delta w/ POD w/o POD w/ POD I w/o POD RC 2
High TASCS, TT Medium 3
4 1
-1.51 E-08
-2.80E-09
-5.94E-10
-1.10E-10 RC 2
High TASCS Medium 2
2 0
-6.72E-09 no change
-2.64E-10 no change RC 2
High TT Medium 3
3 0
-1.01 E-08 no change
-3.96E-10 no change RC 4
High None Low 31 32 1
-1.40E-10
-1.40E-10
-5.50E-12
-5.50E-12 RC Total
-3.21 E-08
-2.94E-09
-1.26E-09
-1.16E-10 AC 2
High TT, IGSCC Medium 1
1 0
-3.36E-09 no change
-1.32E-10 no change AC 2
High TASCS Medium 0
0 0
no change no change no change no change AC 2
High IGSCC Medium 0
0 0
no change no change no change no change AC 4
High None Low 24 30 6
-8.40E-10
-8.40E-10
-3.30E-11
-3.30E-11 AC 5a Medium IGSCC Medium 2
1
-1 1.00E-11 1.OOE-11 1.00E-12 1.OOE-12 AC 6a Medium None Low 15 0
-15 negligible negligible negligible negligible AC 7a Low None Low 0
0 0
no change no change no change no change AC Total
-4.19E-09
-8.30E-10
-1.64E-10
-3.20E-11 SI 2
High IGSCC Medium 2
1
-1 2.80E-09 2.80E-09 1.10E-10 1.10E-10 SI 4
High None Low 17 20 3
-4.20E-10
-4.20E-10
-1.65E-11
-1.65E-11 SI 6a Medium None Low 3
0
-3 negligible negligible negligible negligible SI 7a Low None Low 0
0 0
no change no change no change no change SI Total 2.38E-09 2.38E-09 9.35E-11 9.35E-11 CVC 4
High None Low 0
4 4
-5.60E-10
-5.60E-10
-2.20E-11
-2.20E-11 CVC 6a Medium None Low 0
0 0
no change no change no change no change CVC Total
-5.60E-10
-5.60E-10
-2.20E-11I
-2.20E-11 Page 23 of 34
Table 3.6-1 Unit I - Risk Impact Analysis Results Consequence Failure Potential Inspections CDF lmpact(3)
LERF Impacte3)
Rank DMs Rank Section XI(21 RI-ISI Delta wl POD w/o POD w/ POD [ w/o POD MS 4
High None Low 14 8
-6 8.40E-10 8.40E-10 3.30E-11 3.30E-11 MS Total 8.40E-10 8.40E-10 3.30E-11 3.30E-11 FW 2(1)
High TASCS, (FAC)
Medium (High) 4 3
-1
-8.40E-09 2.80E-09
-3.30E-10 1.10E-10 FW 2
High TASCS Medium 1
1 0
-3.36E-09 no change
-1.32E-10 no change FW 6a (3)
Medium None (FAC)
Low (High) 2 0
-2 negligible negligible negligible negligible FW 6a Medium None Low 6
0
-6 negligible negligible negligible negligible FW Total
-1.18E-08 2.80E-09
-4.62E-10 1.10E-10 AFW(4) 2 High TASCS Medium 1
2 1
-8.40E-09
-2.80E-09
-3.30E-10
-1.10E-10 AFW(4) 4 High None Low 15 16 1
-1.40E-10
-1.40E-10
-5.50E-12
-5.50E-12 AFW Total
-8.54E-09
-2.94E-09
-3.36E-10
-1.16E-10 Grand Total
-5.39E-08
-1.25E-09
-2.12E-09
-4.85E-11 Notes
- 1. Systems are described in Table 3.1-1.
- 2.
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-1 12657.
- 3.
Per Section 3.7.1 of EPRI TR-1 12657, the contribution of low risk categories 6 and 7 need not be considered in assessing the change in risk. Hence, the word "negligible" is given in these cases in lieu of values for CDF and LERF Impact. In those cases where no inspections were being performed previously via Section XI, and none are planned for RI-ISI purposes, "no change" is listed instead of "negligible".
- 4.
The 1986 Edition of ASME Section XI did not require examinations on the AFW piping welds listed in this table. However, in accordance with the 1998 Edition through 2000 Addenda of ASME Section XI, AFW piping is being added to the ISI Program for the fourth ISI interval. The criteria of the 1998 Edition through 2000 Addenda of ASME Section XI were used to project the number of AFW welds that would have required examination during the fourth ISI interval per a standard ISI Program application.
Page 24 of 34
Table 3.6-2 Unit 2 - Risk Impact Analysis Results System(1)
Category Consequence Failure Potential Inspections CDF Impact(3)
LERF Impact(3)
Rank DIs Rank Section XI(2)
RI-ISI Delta w/ POD w/o POD w/ POD w/o POD RC 2
High TASCS, TT Medium 3
3 0
-1.01 E-08 no change
-3.96E-10 no change RC 2
High TASCS Medium 1
2 1
-8.40E-09
-2.80E-09
-3.30E-10
-1.10E-10 RC 2
High TT Medium 3
2
-1
-5.04E-09 2.80E-09
-1.98E-10 1.10E-10 RC 2
High PWSCC Medium 2
1
-1 2.80E-09 2.80E-09 1.10E-10 1.10E-10 RC 4
High None Low 31 28
-3 4.20E-10 4.20E-10 1.65E-11 1.65E-11 RC Total
-2.03E-08 3.22E-09
-7.98E-10 1.27E-10 AC 2
High TT, IGSCC Medium 0
1 1
-5.04E-09
-2.80E-09
-1.98E-10
-1.10E-10 AC 2
High TASCS Medium 1
0
-1 1.68E-09 2.80E-09 6.60E-11 1.10E-10 AC 2
High IGSCC Medium 0
0 0
no change no change no change no change AC 4
High None Low 12 28 16
-2.24E-09
-2.24E-09
-8.80E-11
-8.80E-11 AC 5a Medium IGSCC Medium 0
1 1
-1.00E-11
-1.o00E-11
-1.00E-12
-1.00E-12 AC 6a Medium None Low 14 0
-14 negligible negligible negligible negligible AC 7a Low None Low 0
0 0
no change no change no change no change AC Total
-5.61E-09
-2.25E-09
-2.21E-10
-8.90E-11 SI 2
High IGSCC Medium 0
1 1
-2.80E-09
-2.80E-09
-1.10E-10
-1.10E-10 SI 4
High None Low 24 27 3
-4.20E-10
-4.20E-10
-1.65E-11
-1.65E-11 SI 5a Medium IGSCC Medium 0
1 1
-1.OOE-11
-1.00E-11
-1.00E-12
-1.OOE-12 SI 6a Medium None Low 9
0
-9 negligible negligible negligible negligible SI 7a Low None Low 0
0 0
no change no change no change no change SI Total
-3.23E-09
-3.23E-09
-1.28E-10
-1.28E-10 CVC 4
High None Low 0
4 4
-5.60E-10
-5.60E-10
-2.20E-11
-2.20E-11 CVC 5a Medium TT Medium 0
1 1
-1.80E-11
-1.OOE-11
-1.80E-12
-1.00E-12 CVC 6a Medium None Low 0
0 0
no change no change no change no change CVC Total
-5.78E-10
-5.70E-10
-2.38E-11
-2.30E-11 Page 25 of 34
Table 3.6-2 System~')
Category Consequence Failure Potential Inspections CDF Impactf3)
LERF Impactf3)
Rank DMs Rank Section XI(2)
RI-ISI Delta w/ POD wlo POD w/ POD I w/o POD MS 4
High None Low 9
9 0
no change no change no change no change MS Total no change no change no change no change FW 2(1)
High TASCS, (FAC)
Medium (High) 3 2
-1
-5.04E-09 2.80E-09
-1.98E-10 1.10E-10 FW 2
High TASCS Medium 0
1 1
-5.04E-09
-2.80E-09
-1.98E-10
-1.10E-10 FW 6a (3)
Medium None (FAC)
Low (High) 3 0
-3 negligible negligible negligible negligible FW 6a Medium None Low 12 0
-12 negligible negligible negligible negligible FWTotal
-1.01E-08 no change
-3.96E-10 no change AFW(4) 2 High TASCS Medium 1
2 1
-8.40E-09
-2.80E-09
-3.30E-10
-1.10E-10 AFW(4) 4 High None Low 15 18 3
-4.20E-10
-4.20E-10
-1.65E-11
-1.65E-11 AFW Total
-8.82E-09
-3.22E-09
-3.47E-10
-1.27E-10 Grand Total
-4.86E-08
-6.05E-09
-1.91 E-09
-2.40E-10 Notes
- 1. Systems are described in Table 3.1-2.
- 2.
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-1 12657.
- 3.
Per Section 3.7.1 of EPRI TR-1 12657, the contribution of low risk categories 6 and 7 need not be considered in assessing the change in risk. Hence, the word "negligible" is given in these cases in lieu of values for CDF and LERF Impact. In those cases where no inspections were being performed previously via Section XI, and none are planned for RI-ISI purposes, "no change" is listed instead of "negligible".
- 4.
The 1986 Edition of ASME Section XI did not require examinations on the AFW piping welds listed in this table. However, in accordance with the 1998 Edition through 2000 Addenda of ASME Section XI, AFW piping is being added to the ISI Program for the fourth ISI interval. The criteria of the 1998 Edition through 2000 Addenda of ASME Section Xl were used to project the number of AFW welds that would have required examination during the fourth ISI interval per a standard ISI Program application.
Page 26 of 34 Unit 2 - Risk Impact Analysis Results
Table 5-1-1 Unit I - Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Region High Risk Region Medium Risk Region Low Risk Region Systm~l)
Code I2X2 Category Weld 1986 Section XI(') EPRI TR-112657 Weld 1986 Section XI(2 ) EPRI TR-112657 Weld 1986 Section XI( 21 EPRI TR-112657 Count Vol/Sur Sur Only RI-ISI Other(3 Count Vol/Sur Sur Only RI-ISI IOther(3)
Count Vol/Sur Sur Only RI-ISI IOther(3)
B-F 1
1 0
0 14 14 0
2 RC B-J 32 7
4 9
303 17 69 30 B-J 2
1 0
1 25 5
1 11 21 7
0 0
AC C-F-1 2
0 0
0 276 21 0
20 98 8
0 0
B-J 2
2 0
1 18 2
0 12 105 0
39 0
SI C-F-i 180 15 12 8
220 3
11 0
CVC B-J 32 0
7 4
199 0
49 0
MS C-F-2 80 14 2
8 FW C-F-2 11 5
0 4(4) 37 8
0 0
AFW(5)
C-F-2 6
1 0
2 158 15 0
16 B-F 1
1 0
0 14 14 0
2 B-J 36 10 4
11 378 24 77 57 325 7
88 0
Total C-F-1 2
0 0
0 456 36 12 28 318 11 11 0
C-F-2 17 6
0 6
238 29 2
24 37 8
0 0
Notes I.
Systems are described in Table 3.1-1.
- 2.
Since no examination selections had been made for the fourth interval ISI Program prior to the development of the RI-ISI Program, the third interval selections were used for comparison purposes. The Code of record for the third interval was the 1986 Edition of ASME Section Xl. The Code Categories listed in the table are therefore in accordance with the 1986 Edition of ASME Section Xl.
- 3.
The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-112657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, PBNP Unit 1 achieved a 9.3% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.
Page 27 of 34
Notes for Table 5-1-1 (cont'd)
- 4.
Three of the four welds were selected for examination by both the FAC and RI-ISI Programs. Since a damage mechanism other than FAC was identified, these welds will be subject to both FAC and RI-ISI examinations.
- 5.
The 1986 Edition of ASME Section X1 did not require examinations on the AFW piping welds listed in this table. However, in accordance with the 1998 Edition through 2000 Addenda of ASME Section Xl, AFW piping is being added to the ISI Program for the fourth ISI interval. The criteria of the 1998 Edition through 2000 Addenda of ASME Section Xl were used to project the number of AFW welds that would have required examination during the fourth ISI interval per a standard ISI Program application.
Page 28 of 34
Table 5-1-2 Unit 2 - Inspection Location Selection Comparison Between 1986 Section XI Code and EPRI TR-1 12657 by Risk Region High Risk Region Medium Risk Region Low Risk Region Systm~l)
Code2X12X2 Category Weld 1986 Section XIl 2) EPRI TR-112657 Weld 1986 Section Xl(21 EPRI TR-112657 Weld 1986 Section XIt21 EPRI TR-112657 Count Vol/Sur Sur Only RI-ISI JOther(3)
Count Vol/Sur Sur Only RI-ISI Other)
Count Vol/Sur SurOnly RI-ISI Other(3)
B-F 3
3 0
1 12 12 0
0 RC B-J 29 6
1 7
265 19 63 28 B-J 2
0 0
1 19 2
1 9
15 3
0 0
AC C-F-1 2
1 0
0 257 10 0
20 93 11 0
0 B-J 2
0 0
1 23 6
0 11 80 4
17 0
SI C-F-1 248 18 11 17 224 5
9 0
CVC B-J 37 0
7 5
134 0
44 0
MS C-F-2 89 9
3 9
FW C-F-2 8
3 0
3(4) 44 15 0
0 AFW(5)
C-F-2 8
1 0
2 179 15 0
18 B-F 3
3 0
1 12 12 0
0 B-J 33 6
1 9
344 27 71 53 229 7
61 0
Total C-F-1 2
1 0
0 505 28 11 37 317 16 9
0 C-F-2 16 4
0 5
268 24 3
27 44 15 0
0 Notes
- 1. Systems are described in Table 3.1-2.
- 2.
Since no examination selections had been made for the fourth interval ISI Program prior to the development of the RI-ISI Program, the third interval selections were used for comparison purposes. The Code of record for the third interval was the 1986 Edition of ASME Section X1. The Code Categories listed in the table are therefore in accordance with the 1986 Edition of ASME Section Xl.
- 3.
The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, PBNP Unit 2 achieved greater than a 10% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.
Page 29 of 34
Notes for Table 5-1-2 (cont'd)
- 4.
Two of the three welds were selected for examination by both the FAC and RI-ISI Programs. Since a damage mechanism other than FAC was identified, these welds will be subject to both FAC and RI-ISI examinations.
- 5.
The 1986 Edition of ASME Section XI did not require examinations on the AFW piping welds listed in this table. However, in accordance with the 1998 Edition through 2000 Addenda of ASME Section XI, AFW piping is being added to the ISI Program for the fourth ISI interval. The criteria of the 1998 Edition through 2000 Addenda of ASME Section Xl were used to project the number of AFW welds that would have required examination during the fourth ISI interval per a standard ISI Program application.
Page 30 of 34
Table 5-2-1 Unit I - Inspection Location Selection Comparison Between 1986 Section XI Code and EPRI TR-112657 by Risk Category System(1)
Risk Consequence Failure Potential Code Weld 1986 Section XI(2) EPRI TR-112657 Category Rank Rank DMs J
Rank Category Count Vol/Sur Sur Only RI-ISI [Other(3)
RC 2
High High TASCS, TT Medium B-J 10 3
0 4
RC 2
High High TASCS Medium B-J 6
2 1
2 B-F 1
1 0
0 RC 2
High High TT Medium B-J 16 2
3 3
B-F 14 14 0
2 RC 4
Medium High None Low B-J 303 17 69 30 AC 2
High High TT, IGSCC Medium B-J 1
1 0
1 AC 2
High High TASCS Medium C-F-1 2
0 0
0 AC 2
High High IGSCC Medium B-J 1
0 0
0 B-J 21 3
1 10 AC 4
Medium High None Low C-F-i 276 21 0
20 AC 5
Medium Medium IGSCC Medium B-J 4
2 0
1 B-J 21 7
0 0
AC 6
Low Medium None Low C-F-1 94 8
0 0
AC 7
Low Low None Low C-F-1 4
0 0
0 SI 2
High High IGSCC Medium B-J 2
2 0
1 B-J 18 2
0 12 SI 4
Medium High None Low C-F-1 180 15 12 8
B-J 105 0
39 0
SI 6
Low Medium None Low C-F-1 200 3
11 0
SI 7
Low Low None Low C-F-1 20 0
0 0
CVC 4
Medium High None Low B-J 32 0
7 4
CVC 6
Low Medium None Low B-J 199 0
49 0
MS 4
Medium High None Low C-F-2 80 14 2
8 Page 31 of 34
Table 5-2-1 (cont'd)
Unit I - Inspection Location Selection Comparison Between 1986 Section XI Code and EPRI TR-112657 by Risk Category System(l)
Risk Consequence Failure Potential Code Weld 1986 Section Xl(2) EPRI TR-112657 Category Rank Rank DMs Rank Category Count Vol/Sur Sur Only RI-ISI I Other(3)
FW 2 (1)
High (High)
High TASCS, (FAC)
Medium (High)
C-F-2 9
4 0
3(4)
FW 2
High High TASCS Medium C-F-2 2
1 0
1 FVV 6 (3)
Low (High)
Medium None (FAC)
Low (High)
C-F-2 10 2
0 0
FW 6
Low Medium None Low C-F-2 27 6
0 0
AFW(5) 2 High High TASCS Medium C-F-2 6
1 0
2 AFWV*)
4 Medium High None Low C-F-2 158 15 0
16 Notes
- 1. Systems are described in Table 3.1-1.
- 2.
Since no examination selections had been made for the fourth interval ISI Program prior to the development of the RI-ISI Program, the third interval selections were used for comparison purposes. The Code of record for the third interval was the 1986 Edition of ASME Section Xl. The Code Categories listed in the table are therefore in accordance with the 1986 Edition of ASME Section Xl.
- 3.
The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, PBNP Unit 1 achieved a 9.3% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.
- 4. These three welds were selected for examination by both the FAC and RI-ISI Programs. Since a damage mechanism other than FAC was identified, these welds will be subject to both FAC and RI-ISI examinations.
- 5.
The 1986 Edition of ASME Section Xl did not require examinations on the AFW piping welds listed in this table. However, in accordance with the 1998 Edition through 2000 Addenda of ASME Section XI, AFW piping is being added to the ISI Program for the fourth ISI interval. The criteria of the 1998 Edition through 2000 Addenda of ASME Section XI were used to project the number of AFW welds that would have required examination during the fourth ISI interval per a standard ISI Program application.
Page 32 of 34
Table 5-2-2 Unit 2 - Inspection Location Selection Comparison Between 1986 Section Xl Code and EPRI TR-112657 by Risk Category System(l)
Risk Consequence Failure Potential Code Weld 1986 Section XI(2) EPRI TR-112657 Category Rank Rank DMs Rank Category Count Vol/Sur Sur Only RI-ISI ]
Other(3)
RC 2
High High TASCS, TT Medium B-J 10 3
0 3
RC 2
High High TASCS Medium B-J 6
1 0
2 B-F 1
1 0
0 RC 2
High High TT Medium B-J 13 2
1 2
RC 2
High High PWSCC Medium B-F 2
2 0
1 B-F 12 12 0
0 RC 4
Medium High None Low B-J 265 19 63 28 AC 2
High High TT, IGSCC Medium B-J 1
0 0
1 AC 2
High High TASCS Medium C-F-1 2
1 0
0 AC 2
High High IGSCC Medium B-J 1
0 0
0 B-J 17 2
1 8
AC 4
Medium High None Low C-F-I 257 10 0
20 AC 5
Medium Medium IGSCC Medium B-J 2
0 0
1 B-J 15 3
0 0
AC 6
Low Medium None Low C-F-1 89 11 0
0 AC 7
Low Low None Low C-F-1 4
0 0
0 SI 2
High High IGSCC Medium B-J 2
0 0
1 B-J 22 6
0 10 SI 4
Medium High None Low C-F-1 248 18 11 17 SI 5
Medium Medium IGSCC Medium B-J 1
0 0
1 B-J 80 4
17 0
SI 6
Low Medium None Low C-F-I 199 5
6 0
SI 7
Low Low None Low C-F-I 25 0
3 0
Page 33 of 34
Table 5-2-2 (cont'd)
Unit 2 - Inspection Location Selection Comparison Between 1986 Section XI Code and EPRI TR-112657 by Risk Category System(1)
Risk Consequence Failure Potential Code Weld 1986 Section Xl(z) EPRI TR-112657 Category Rank Rank DMs
]
Rank Category Count Vol/Sur SurOnly RI-ISI Other(3 )
CVC 4
Medium High None Low B-J 36 0
7 4
CVC 5
Medium Medium TT Medium B-J 1
0 0
1 CVC 6
Low Medium None Low B-J 134 0
44 0
MS 4
Medium High None Low C-F-2 89 9
3 9
FW 2 (1)
High (High)
High TASCS (FAC)
Medium (High)
C-F-2 5
3 0
2(4)
FW 2
High High TASCS Medium C-F-2 3
0 0
1 FW 6 (3)
Low (High)
Medium None (FAC)
Low (High)
C-F-2 6
3 0
0 FW 6
Low Medium None Low C-F-2 38 12 0
0 AFW(5) 2 High High TASCS Medium C-F-2 8
1 0
2 AFW(5) 4 Medium High None Low C-F-2 179 15 0
18 Notes
- 1.
Systems are described in Table 3.1-2.
- 2.
Since no examination selections had been made for the fourth interval ISI Program prior to the development of the RI-ISI Program, the third interval selections were used for comparison purposes. The Code of record for the third interval was the 1986 Edition of ASME Section X1. The Code Categories listed in the table are therefore in accordance with the 1986 Edition of ASME Section XI.
- 3.
The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, PBNP Unit 2 achieved greater than a 10% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.
- 4.
These two welds were selected for examination by both the FAC and RI-ISI Programs. Since a damage mechanism other than FAC was identified, these welds will be subject to both FAC and RI-ISI examinations.
- 5.
The 1986 Edition of ASME Section XI did not require examinations on the AFW piping welds listed in this table. However, in accordance with the 1998 Edition through 2000 Addenda of ASME Section X1, AFW piping is being added to the ISI Program for the fourth ISI interval. The criteria of the 1998 Edition through 2000 Addenda of ASME Section XI were used to project the number of AFW welds that would have required examination during the fourth ISI interval per a standard ISI Program application.
Page 34 of 34