RBG-46922, Request for Alternative - Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716

Request for Alternative - Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716

ML091740306
Person / Time
Site:	River Bend
Issue date:	06/16/2009
From:	Roberts J Entergy Corp, Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
N-716, RBG-46922
Download: ML091740306 (73)

Text

fRiver Bend Station En t LI&Igy 5485 U.S. Highway 61N St. Francisville, LA 70775 Tel 225-381-4149 Jerry C. Roberts Director, Nuclear Safety Assurance RBG-46922 June 16, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Request for Alternative - Implementation of a Risk-Informed Inservice Inspection Program Based on ASME Code Case N-716 River Bend Station, Unit 1 Docket No. 50-458 License No. NPF-47

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Entergy Operations, Inc. (Entergy) hereby requests authorization to implement a risk-informed Inservice Inspection (RI-ISI) program based on the American Society of Mechanical Engineers (ASME) Code Case N-716, as documented in the attached Request for Alternative RBS-ISI-013. RBS-ISI-013 is being submitted in a template format as Attachment 1. This template format is similar to the submittals the NRC Staff has approved for Waterford 3 and Grand Gulf. This format is also similar to the recently submitted request for alternative by Calvert Cliffs and Arkansas Nuclear One for the same subject. This request includes information to address NRC requests for additional information available at the time of development of this submittal.

In accordance with 10 CFR 50.55a(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. Entergy believes the proposed alternative meets this requirement.

Entergy requests to implement this alternative beginning with the third period of the second ISI interval to cover the remaining weld examinations that were not performed during the second interval. The second ISI interval was previously extended by RBS-ISI-005 as approved by NRC on May 17, 2007 (TAC No. MD3442). A separate relief request (RBS-ISI-012) has been submitted requesting further extension of the second interval to allow for completion of NRC's review of RBS-ISI-013.

Entergy requests approval of the proposed alternative by December 1, 2010. RBS will withdraw the Request for Alternative CEP-ISI-007 pertaining to the application of Code Case N-663 for use at RBS upon NRC approval of this RI ISI program submittal. Although this request is neither exigent nor emergency, your prompt review is requested.

The request for alternative includes several new commitments that are summarized in . K)

RBG-46922 Page 2 of 2 If you have any questions or require additional information, please contact David Lorfing, Manager, Licensing at (225) 381-4157.

Sincerely,-

teor, Nuclear Safety Assurance River Bend Station - Unit 1 JCR/DNL/bmb Attachments:

1. Request for Alternative RBS-lSl-01 3

2. Licensee Identified Commitments cc: Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125

/

NRC Senior Resident Inspector P. O. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS 0-7 D1 Washington, DC 20555-0001 Mr. Jeffrey P. Meyers Louisiana Department of Environmental Quality Office of Environmental Compliance Attn. OEC - ERSD P. O. Box 4312 Baton Rouge, LA 70821-4312

ATTACHMENT I TO RBG-46922 REQUEST FOR ALTERNATIVE RBS-ISI-013

REQUEST FOR ALTERNATIVE ENTERGY OPERATIONS, INC.

RIVER BEND STATION - UNIT I REQUEST FOR ALTERNATIVE RBS-ISI-013 APPLICATION OF ASME CODE CASE N-716 RISK-INFORMED / SAFETY-BASED INSERVICE INSPECTION PROGRAM PLAN Table of Contents 1 Introduction 2 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 2 1.2 PRA Quality 2

2. Proposed Alternative to Current Inservice Inspection Programs 3 2.1 ASME Section XI 3 2.2 Augmented Programs 3

3. Risk-Informed / Safety-Based ISI Process 4 3.1 Safety Significance Determination 4 3.2 Failure Potential Assessment 5 3.3 Element and NDE Selection 7 3.3.1 Additional Examinations 8 3.3.2 Program Relief Requests 8 3.4 Risk Impact Assessment 9 3.4.1 Quantitative Analysis 9 3.4.2 Defense-in-Depth 12

4. Implementation and Monitoring Program 13

5. Proposed ISI Program Plan Change 14

6. References/Documentation 14 Appendix 1 PRA Model Capability for Use in RI-ISI Applications 15 to RBG-46922 Page 2 of 68 ENTERGY OPERATIONS, INC.

RIVER BEND STATION - UNIT I REQUEST FOR ALTERNATIVE RBS-ISI-01 3

1. INTRODUCTION River Bend Station - Unit 1 (RBS) is currently in the second 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. RBS plans to complete the current (second) ISI interval by implementing a risk-informed / safety-based inservice inspection (RISB) program during the third inspection period of the interval. Entergy will also implement 100% of the RIS_B program in the third ISI interval.

The ASME Section Xl code of record for the second ISI interval at RBS is the 1992 Edition for Examination Category B-F, B-J, C-F-i, and C-F-2 Class 1 and 2 piping components. The ASME Section XI code of record for the third ISI interval- at RBS is the 2001 Edition through 2003 Addenda for items in these Examination Categories.

The objective of this submittal is to request the use of the RIS_B process for the ISI of Class 1 and 2 piping. The RISB process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements, Section X1 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 ProbabilisticRisk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis, and RG 1.178, An Approach for Plant-Specific Risk-Informed DecisionmakingInservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.

1.2 Probabilistic Safety Assessment (PSA) Quality The River Bend Station PRA has been demonstrated to be adequate for this application. The PRA and its supporting processes are described in further detail in Appendix 1. As described in the Appendix, the RBS PRA internal events model has been reviewed as part of the Boiling Water Reactor Owners Group (BWROG) Peer Review process in 1998. A self-assessment of the current PRA model against Regulatory Guide 1.200 was conducted in Fall 2008. Results of the self-assessment are discussed in the Appendix. The Internal Flooding PRA was updated to fully meet RG 1.200 requirements in 2009. Future PRA changes under the PRA maintenance and update process will address identified gaps against Regulatory Guide 1.200. As discussed in the Appendix, most of the gaps are considered documentation issues and all gaps have been reviewed to support the conclusion that the RBS PRA is fully capable of supporting the request to use the RIS_B process, based upon ASME Code Case N-716, for a Risk-Informed In-Service Inspection program at River Bend.

to RBG-46922 Page 3 of 68

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section Xl ASME Section XI Examination Categories B-F, B-J, C-F-i, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components, except as amended by the application of ASME Code Case N-663 (Request for Alternative CEP-ISI-007) that was approved for use at RBS by the NRC on August 26, 2003.

The alternative RISB 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 a*ccordance 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 RISB 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 (e.g., Class 1 and 2 piping).

The original plant augmented inspection program for high-energy line breaks, implemented in accordance with RBS Final Safety Analysis Report (FSAR) Sections 5.2.4.8 "Augmented Inservice Inspection to Protect Against Postulated Class 1 Piping Failures" and 6.6.8, "Augmented Inservice Inspection to Protect Against Postulated Class 2 Piping Failures,"

were revised in accordance with the risk-informed break exclusion region methodology (RI-BER) described in EPRI Report 1006937, Extension of EPRI Risk Informed IS! Methodology to Break Exclusion Region Programs. EPRI Report 1006937 was approved by the NRC in 2002. The results of the RI-BER application demonstrated that the volumetric examination requirement for this scope of piping could be reduced from 100% to approximately 15%. As a result, 15% of the BER population will be examined during the course of each ten-year interval which exceeds the 10% requirement imposed by Code Case N-716.

The RBS augmented inspection program for intergranular stress corrosion cracking (IGSCC) per Generic Letter (GL) 88-01, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping, is relied upon to manage this damage mechanism. GL 88-01 specifies the examination extent and frequency requirements for austenitic stainless steel welds classified as Categories A through G, depending on their susceptibility to IGSCC. In accordance with EPRI TR-112657, piping welds identified as Category A are considered resistant to IGSCC and are assigned a low failure potential provided no other damage mechanisms are present. Consequently, the examination of welds identified as Category A inspection locations is subsumed by the RIS_B Program.

The existing RBS augmented inspection program for the other piping welds susceptible to IGSCC (Categories "B" and "C") remains unaffected by the RISB Program submittal.

The plant augmented inspection program for flow accelerated corrosion (FAC) per Generic Letter (GL) 89-08, Erosion/Corrosion-InducedPipe Wall Thinning, is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B Program.

Attachment 1 to RBG-46922 Page 4 of 68

3. RISK-INFORMED / SAFETY-BASED ISI PROCESS The process used to develop the RIS_B 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 RISB Program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information including the existing plant ISI 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. HSS welds are determined in accordance with the requirements below. LSS welds include, all other Class 2, 3, or Non-Class welds.

.(1) Class 1 portions of the reactor coolant pressure boundary (RCPB), 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 RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., 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 (i.e.,

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 within the break exclusion region (> NPS 4) 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 (and per NRC feedback on the Grand Gulf and DC Cook RIS_B pilot applications 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.

to RBG-46922 Page 5 of 68 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 EPRI TR-112657 (i.e., the EPRI RI-ISI methodology), with the exception of the deviation discussed below.

Table 3.2 summarizes the failure potential assessment by 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 RBS. 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 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 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 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.

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 to RBG-46922 Page 6 of 68 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., 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 RISB 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 TASCS criteria at several facilities, including Comanche Peak (NRC letter dated September 28, 2001) and South Texas Project (NRC letter dated March 5, 2002).

• Attachment 1 to RBG-46922 Page 7 of 68 3.3 - Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RISB pilot applications provide 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) (i.e., 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 Table 3.3. Section 4 of EPRI TR-112657 was used as guidance in determining the examination requirements for these locations.

Class I Welds( 1) Class 2 Welds (2) Class 3 Welds(3 ) All Piping Welds(4)

UUnit n it ..... . . . .* . . ..

Total SeleCted Total Selected Total Selected Total SeleCted 1 763 83 1338 0 4 0 2105 83 Notes (1) Includes all Category B-F and B-J locations. All 763 Class 1 piping weld locations are HSS.

(2) Includes all Category C-F-1 and C-F-2 locations. Of the 1338 Class 2 piping weld locations, 13 are HSS and the remaining 1325 are LSS.

(3) All four of these Class 3 piping weld locations are HSS.

(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 X1 Program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RISB Program.

to RBG-46922 Page 8 of 68 Prior to developing the RIS_B Program, RBS had planned to inspect locations scheduled for examination under a traditional ASME Section Xl inspection program. Examination activities during refueling outages are planned well in advance. In general, only designated plant areas and components are accessible for examination during a given refueling outage due to other ongoing plant maintenance and modification activities.

Hence, any location previously scheduled for examination in the third period via the traditional program will remain scheduled for examination in the third period, for locations selected for RISB Program purposes. Additional samples will be selected, if necessary, to achieve equal representation of the degradation mechanisms. Other factors, such as accessibility and scaffolding requirements, will also factor into the selection process.

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 Xl, IWB-3500 and/or IWB-3600. As part of performing 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 Xl. 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 Xl Code Cases. The need for extensive root cause analysis beyond that required for IWB-3600 evaluation is dependent on practical considerations (i.e., 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(a)(3) 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 theevaluation to be submitted to the NRC. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section Xl.

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 RISB 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 at the time of the examination fail to meet the >90% coverage requirement, the process outlined in 10 CFR 50.55a will be followed.

to RBG-46922 Page 9 of 68 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 ISI 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, RBS 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 submitted per the guidance of 10 CFR 50.55a(g)(5)(iv) within one (1) year after the end of the interval.

Request for Alternative CEP-ISI-007 pertaining to the application of Code Case N-663 will be withdrawn for use at RBS upon NRC approval of the RISB Program submittal.

3.4 Risk Impact Assessment The RISB 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 remain neutral or decrease when compared to that estimated from current requirements.

This evaluation categorized segments as HSS or LSS in accordance with Code Case N-716, and then determined what inspection changes are 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 RISB 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 Code Case N-716 has adopted the EPRI TR-1 12657 process for risk impact analyses whereby limits are imposed to ensure that the change in risk of implementing the RIS_B Program meets the requirements of RG 1.174 and. 1.178. The EPRI criterion requires that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, respectively.

For LSS welds, CCDP and CLERP values of 1E-4 and 1E-5 are generally conservatively used, unless pipe segments in the plant internal flooding study are found with higher values. For the RBS RISB application, CCDP and CLERP values of 3.4E-4 and 1.4E-5 have been used for LSS welds to bound plant internal flooding study results. The 3.4E-4 and 1.4E-5 values used for CCDP and CLERP is based on results from the plant internal flooding study for a postulated rupture of Class 2 Feedwater system piping outside containment and have been conservatively applied as an upper bound for all LSS welds.

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 to RBG-46922 Page 10 of 68 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.

In order to streamline the risk impact assessment, a review was conducted to verify that the LSS piping was not susceptible to FAC or water hammer. This review was conducted similar to that done for a traditional RI-ISI application. Thus, the high failure potential category is not applicable to LSS piping. In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g., to determine if thermal fatigue is applicable), these locations were conservatively assigned to the Medium failure potential ("Assume Medium" in Table 3.4-1) for use in the change-in-risk assessment. Experience with previous industry RI-ISI applications shows this to be conservative.x-RBS has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716 that is consistent with the "Simplified Risk Quantification Method" described in Section 3.7 of EPRI TR-1 12657. The analysis estimates the net change in risk due to the positive and negative influences of adding and removing locations from the inspection program.

The CCDP and CLERP values used to assess risk impact were estimated based on pipe break location. Based on these estimated values, a corresponding consequence rank was assigned per the requirements of EPRI TR-112657 and upper bound threshold values were used as provided below. Consistent with the EPRI risk-informed methodology, the upper bound for all HSS break-locations that fall within the high consequence rank range was based on the highest CCDP value obtained dependent upon whether the piping break occurs inside or outside of containment. For piping breaks inside containment, the upper bound is based on RBS plant specific Initiator T3B1 (i.e., loss of Feedwater, condenser, reactor core isolation cooling and shutdown cooling). For piping breaks outside containment, the upper bound is based on RBS plant-specific flood scenarios F-4-2-1-21a and F-4-2-1-21b that assess a rupture of Class 1 HSS and Class 2 LSS Feedwater system piping outside containment.

to RBG-46922 Page 11 of 68 CCDP and CLERP Values Based on Break Location Break Location Estimated Consequence Upper Bound Designation CCDP CLERP Rank CCDP CI..ERP LOCA 1.OE-4 2.3E-6 HIGH 1.7E-4 3.5E-5 Pipe breaks that result in a LOCA - Estimated based on highest CCDP for LOCA (Intermediate Break) from PSA model ILOCA <1.OE-6 <1.OE-7 MEDIUM(1 ) 1.OE-4 1.OE-5 Pipe breaks that result in an isolable LOCA inside containment - Estimated based on Intermediate LOCA CCDP of 1.OE-4 and valve fail to close probability of 3.4E-3 ILOCA - FW 1.7E-4 3.5E-5 HIGH 1.7E-4 3.5E-5 Pipe breaks that result in an isolable LOCA inside containment - Estimated based on loss of Feedwater, condenser, reactor core isolation cooling and shutdown cooling ILOCA - OC 3.4E-4 1.4E-5 HIGH 3.4E-4 1.4E-5 Pipe breaks that result in an isolable LOCA outside containment - Estimated based on flood scenarios F-4-2-1-21a and F-4-2-1-21b that assess a rupture of Class 1 HSS Feedwater system piping outside containment; conservatively applied to all ILOCA - OC designated break locations PLOCA <1.OE-6 <1.OE-7 MEDIUM(1 ) 1.OE-4 1.OE-5 Pipe breaks that result in a potential LOCA - Estimated based on Intermediate LOCA CCDP of 1.OE-4 and valve rupture probability of 1.OE-3 ISLOCA 1.OE-3 1.OE-3 HIGH 1.OE-3 1.OE-3 Pipe breaks that result in an interfacing system LOCA outside containment - Estimated based on valve rupture probability of 1.OE-3 21SLOCA 8.3E-6 8.3E-6 MEDIUM 1.OE-4 1.OE-5 Pipe breaks that result in an interfacing system LOCA outside containment - Estimated based on two valve rupture probability of 8.3E-6 MSD -3 3.4E-6 3.4E&6 MEDIUM 1.OE-4 1.OE-5 Pipe breaks that occur in main steam drain system piping outside containment - Estimated based on an assumed steam LOCA CCDP outside containment of 1.OE-3 and valve fail to close probability of 3.4E-3 DTM - I/MSI - 1 2.OE'6 2.OE-6 MEDIUM 1.OE-4 1.OE-5 Pipe breaks that occur in main steam drain and main steam isolation valve leakage control system piping outside containment - Estimated based on an assumed steam LOCA CCDP outside containment of 1.OE-3 and valve fail to close probability of 2.OE-3 Class 2 LSS 3.4E-4 1.4E-5 HIGH 3.4E-4 1.4E-5 Pipe breaks that occur in Class 2 system piping designated as LSS - Estimated based on flood scenarios F-4-2-1-21a and F-4-2-1-21b that assess a rupture of Class 2 LSS Feedwater system piping outside containment; conservatively applied to all Class LSS system piping Note (1) Although the estimated CCDP and CLERP values for ILOCA and PLOCA break locations fall in the "Low" consequence rank range, a "Medium" consequence rank is conservatively used for risk impact.

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 20x,. 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 to RBG-46922 Page 12 of 68 inspection effectiveness due to an increased POD from application of the RIS_B approach.

Table 3.4-1 presents a summary of the RIS_B Program versus 1992 ASME Section Xl Code Edition program requirements on a "per system" basis. The presence of FAC and IGSCC was adjusted for in the quantitative analysis by excluding their impact on the failure potential rank. The exclusion of the impact of FAC and IGSCC on the failure potential rank and therefore in the determination of the change in risk is appropriate, because FAC and IGSCC are damage mechanisms managed by separate, independent plant augmented inspection programs. The RISB Program credits and relies upon these plant augmented inspection programs to manage these damage mechanisms.

The plant FAC and IGSCC Programs will continue to determine where and when examinations are performed. Hence, since the number of FAC and IGSCC examination locations remains the same "before" and "after" and no delta exist, there is no need to include the impact of FAC and IGSCC in the performance of the risk impact analysis.

As indicated in the following table, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RISB Program, and satisfies the acceptance criteria of RG 1.174 and Code Case N-716.

RBS Risk Impact Results Results ARLERF Results System(1 ) ARCDF w/ POD w/o POD w/ POD w/o POD RPV 2.04E-11 8.84E-11 4.20E-12 1.82E-11 RDS 1.36E-10 1.36E-10 5.60E-12 5.60E-12 RCS -1.02E-11 -1.02E-11 -2.10E-12 -2.10E-12 FWS 1.97E-10 5.44E-10 3.50E-11 9.52E-11 MSS 7.94E-11 7.94E-11 2.96E-12 2.96E-12 SLS -4.90E-12 -4.90E-12 -8.50E-13 -8.50E-13 CSH 4.43E-10 4.77E-10 1.84E-11 2.54E-11 RHS 2.46E-09 2.46E-09 1.16E-10 1.16E-10 CSL 3.09E-10 3.09E-10 1.31 E-11 1.31 E-11 MSI -2.50E-12 -2.50E-12 -2.50E-13 -2.50E-13 lCS 3.39E-10 3.39E-10 1.41E-11 1.41 E-11 WCS 1.OOE-12 1.00E-12 1.OOE-13 1.OOE-13 DTM -5.05E-12 -5.05E-12 -7.75E-13 -7.75E-13 TOTAL 3.97E-09 4.42E-09 2.05E-1O0 -2.86E-10 Note (1) Systems are described in Table 3.1.

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

Attachment 1 to RBG-46922 Page 13 of 68 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 ingredients 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 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, 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 BER break. Finally, Code Case N-716 requires that any plant-specific piping with a contribution to CDF of greater than 1 E-06 (or 1E-07 for LERF) be included in the scope of the application. No such piping was identified at RBS.

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 RISB 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 implemented into the second ISI 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 RISB 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 RISB 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.

to RBG-46922 Page 14 of 68 For preservice examinations, RBS 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. Welds classified as LSS do not require preservice inspection.

5. PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the RISB Program and ASME Section Xl 1992 Code Edition program requirements for in-scope piping is provided in Table 5.

Currently, RBS is in its extended second ISI interval. By letter dated November 1, 2006, the licensee stated that it planned to implement a risk-informed/safety based ISI (RISB) program during the third inspection period of the current (second) ISI interval. In the subject extension request, RBS committed to perform the same percentage of examinations which remained incomplete from the second ISI interval. These welds would be selected from the welds included under the new risk informed program and would have been completed by the end of RF-1 5 currently scheduled for the Fall 2009.

Due to delays in completing the updated flooding study in support of the RIS_B submittal, an additional extension of the second ISI interval is being requested under Request for Alternative RBS-ISI-012. In this request, RBS commits to complete approximately 60% of the remaining examinations selected under the conventional ISI program by the end of RF-15. All of the required first period examinations of the third interval would be performed during RF-16 currently scheduled for 2011.

The third ISI interval will implement 100% of the inspection 'locations selected for-examination per the RIS_B Program. Examinations shall be performed such that the period percentage

'requirements of ASME Section XI are met.

6. REFERENCES/DOCUMENTATION USNRC Safety Evaluation pertaining to the use of ASME Code Case N-663, dated August 26, 2003 (Letter CNRI-2003-00010)

EPRI TR-1 12657, Revised Risk-Informed Inservice Inspection Evaluation Procedure,Rev. B-A EPRI TR-1 018427, Nondestructive Evaluation: ProbabilisticRisk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs ASME Code Case N-716, Alternative Piping Classification and Examination Requirements, Section X1 Division I Regulatory Guide 1. 174, An Approach for Using ProbabilisticRisk 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 ISI based on ASME Code Case N-716, dated September 21, 2007 to RBG-46922 Page 15 of 68 USNRC Safety Evaluation for DC Cook Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based ISI program for Class 1 and 2 Piping Welds, dated September 28, 2007 Supporting Onsite Documentation Structural Integrity (SI) Calculations

• RBS-12Q-301, Rev. 0, DegradationMechanism Evaluation

• RBS-1 2Q-302, Rev. 0, Risk Informed Break Exclusion Region Evaluation for River Bend Station

" RBS-1 2Q-303, Rev. 0, Service History Review

• ENTP-1 9Q-31 0, Rev. 2, DegradationMechanism Evaluation for River Bend

• ENTP-19Q-31 1, Rev. 0, N-716 Evaluation of River Bend Station to RBG-46922 Page 16 of 68 Table 3.1-N-716 Safety Significance Determination N-716 Safety Significance Determination Safety System Description S Weld Wed.ountc_ _ 0 Significance Count W:>E-CD RCPB SDC PWR:F BER >1E-7LERF High Low RPV - Reactor Pressure Vessel (050) 40 _"___"

RDS - Control Rod Drive (052) 48 ,/

RCS - Reactor Coolant (053) 1 V / V 113 V V FWS - Feedwater (107) 10 V / V V 50 V V V 4 / V V 2 / V 23 V MSS - Main Steam (109) 20 V V V 122 V V 16 V SLS - Standby Liquid Control (201) 63 V V CSH - High Pressure Core Spray (203) 19 V V 129 V RHS - Residual Heat Removal (204) 13 V V V 53 V V 866 V CSL - Low Pressure Core Spray (205) 18 V V 78 V MSI - Main Steam Leakage Control (208) 43 V V ICS - Reactor Core Isolation Cooling (209) 9 V V V 10 V V 165 V WCS - Reactor Water Cleanup (601) 8 V V V 83 V V 17 V V DTM --Steam Drains (609) 82 V V

SUMMARY

RESULTS FOR ALL SYSTEMS 10 V V V V 64 V V V

'41 V V V 648 V V 17 V V 1325 V TOTALS 2105 780 1325 to RBG-46922 Page 17 of 68 Table 3.2 Failure Potential Assessment Summary System~1 ) Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS F TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RPV "

RDS(21 RCS V V FWS(2)

MSS(2)

SLS CSH(21 2

1 RHS(

CSL (21 MSI lOS(2)

WCS "

DTM Notes (1) Systems are described in Table 3.1.

(2) A~degradation mechanism assessment was not performed on low safety significant piping segments. This includes the RDS system in its entirety, as well as portions of the FWS, MSS, CSH, RHS, CSL and ICS systems.

to RBG-46922 Page 18 of 68 Table 3.3 N-716 Element Selections System(1 ) Selections HSS DMs(2) RCPB 3 RCPBIFIV( ) RCPBOc BER RPV Required 4 of 40 TASCS, TT, (IGSCC) 4 of 40 3 n/a n/a TASCS,TT 4(7) of TT, (IGSCC) 28 None (IGSCC)

Made 4 TASCS, TT, (IGSCC) 1 4 4 n/a n/a TASCS,TT 1 TT, (IGSCC) 0 None (IGSCC) 2 RDS Required n/a n/a n/a n/a n/a n/a Made n/a n/a n/a n/a n/a n/a RCS Required 12 of 114 n/a 12 of 114 8 n/a n/a Made 12 n/a 12 12 n/a n/a FWS Required 7 of 66 TASCS, TT, (FAC) 7 of 66 5 1 of 10 2 of 14 TAscs, TT 7(17) of 66 TT Made 7 TASCS, TT, (FAC) 5 7 5 2 2 TASCS,TT 2 TT 0 IMSS Required 15 of 142 n/a 15 of 142 10 3 of 25 2 of 20 Made 15 n/a 15 10 3 8,.

SLS Required 7 of 63 n/a 7 of 63 5 3 of 28 n/a Made 7 n/a 7 4 3 n/a CSH Required 2 of 19 TT 1 of 4 2 of 19 2 1 of 4 n/a Made 2 TT 1 2' 1 1 n/a RHS Required 7 of 66 n/a 7 of 66 5 2 of 16 n/a Made 7 n/a 7 5 2 n/a CSL Required 2 of 18 n/a 2 of 18 2 1 of 5 n/a Made 2 n/a 2 1 1 n/a MSI Required 5 of 43 n/a 5 of 43 n/a 5 of 43 n/a Made 5 n/a 5 n/a 5 n/a ICS Required 2 of 19 n/a 2 of 19 2 1 of 5 2 of 9 Made 2 n/a 2 1 1 2 WCS Required 11 of 108 None (FAG) 1 of 4 10 of 91 7 1 of 2 3 of 25 Made 11 None(FAC) 1 11 9 1 3 DTM Required 9 of 82 n/a 9 of 82 6 6 of 52 n/a Made 9 n/a 9 3 6 n/a TOTAL Made 83 13 83 55 25 15 Notes (1) Systems are described in Table 3.1.

(2) For RPV and FWS systems, no more than 10% of HSS piping welds are required to be selected for examination.

(3) For SLS, CSH, CSL, ICS and DTM systems, it was not possible to meet the requirement that 2/3 of the RCPB piping welds selected for examination be located between the first isolation valve and the reactor pressure vessel, while also ensuring that a minimum of 10% of the RCPB piping welds that lie outside containment were selected for examination.

This lesson learned from the RBS RISB application is being addressed in Revision 1 to Code Case N-716.

to RBG-46922 Page 19 of 68 Table 3.4-1 Risk Impact Analysis Results System~1 1 Safety Significancej Break Location(2) DMs Failure Potential Rank(3 J Inspections CDF Impact LERF Impact SXl(4' RISBI-51 Delta w/ POD w/o POD w/IPOD w/o POD RPV High LOCA TASCS, TT, (IGSCC) Medium (Medium) 3 1 -2 0.OOE+00 3.40E-11 0.OOE+00 7.OOE-12 RPV High LOCA TASCS, TT Medium 1 1 0 -2.04E-11 0.OOE+00 -4.20E-12 0.OOE+00 RPV High LOCA TT, (IGSCC) Medium (Medium) 2 0 -2 2.04E-1 1 3.40E-11 4.20E-12 7.OOE-12 RPV High LOCA None (IGSCC) Low (Medium) 22 2 -20 1.70E-11 1.70E-11 3.50E-12 3.50E-12 RPV High LOCA None Low 4 0 -4 3.40E-12 3.40E-12 7.OOE-13 7.00E-13 TOTAL 2.04E-11 8.84E-11 4.20E-12 1.82E-11 RDS Low Class 2 LSS N/A Assume Medium 4 0 -4 1.36E-10 1.36E-10 5.60E-12 5.60E-12 TOTAL 1.36E-10 1.36E-10 5.60E-12 5.60E-12 RCS High LOCA None Low 0 12 12 -1.02E-11 -1.02E-11 -2.10E-12 -2.10E-12 TOTAL I -1.02E-11 -1.02E-1 I -2.10E-12 -2.10E-12 FWS High LOCA TASCS, TT, (FAC) Medium (High) 15 5 -10 0.00E+00 1.70E-10 0.OOE+00 3.50E-11 FWS High LOCA TASCS,TT Medium 16 0 -16 1.63E-i0 2.72E-10 3.36E-11 5.60E-11 FWS High LOCA TT Medium 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00-FWS High ILOCA - FW TASCS, TT Medium 0 0 0 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 FWS High ILOCA- OC TASCS, TT Medium 1 2 1 -1.02E-10 -3.40E-11 -4.20E-12 -1.40E-12 FWS Low Class 2 LSS N/A Assume Medium 4 0 -4 1.36E-10 1.36E-10 5.60E-12 5.60E-12 TOTAL 1.97E-10 5.44E-10 3.50E-11 9.52E-11 MSS High LOCA None Low 8 10 2 -1.70E-12 -1.70E-12 -3.50E-13 -3.50E-13 MSS High ILOCA None Low 1 2 1 -5.OOE-13 -5.00E-13 -5.OOE-14 -5.OOE-14 MSS High ILOCA - OC None Low 11 3 -8 1.36E-11 1.36E-11 5.60E-13 5.60E-13 MSS High PLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 MSS High DTM - 1 / MSI - 1 None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 MSS Low Class 2 LSS N/A Assume Medium 2 0 -2 6.80E-11 6.80E-11 2.80E-12 2.80E-12 TOTAL 7.94E-11 7.94E-11 2.96E-12 2.96E-12 SLS High LOCA None Low 0 4 4 -3.40E-12 -3.40E-12 -7.OOE-13 -7.OOE-13 SLS High PLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 SLS High ISLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 SLS High 21SLOCA None Low 0 3 3 -1.50E-12 -1.50E-12 -1.50E-13 -1.50E-13 TOTAL -4.90E-12 -4.90E-12 -8.50E-1 3 -8.50E-13 to RBG-46922 Page 20 of 68 Table 3.4-1 (Cont'd)

Risk Impact Analysis Results Inspections CDF Impact LERF Impact System~l) Safety Significance Break Location (2)

J DMs Failure Potential Rank(3 ) SXI(4) RIS_B(') Delta w/ POD w/o POD w/ POD w/o POD CSH High LOCA TT Medium 3 1 -2 0.00E+00 3.40E-11 0.OOE+00 7.00E-12 CSH High LOCA None Low 1 0 -1 8.50E-13 8.50E-13 1.75E-13 1.75E-13 CSH High PLOCA None Low 0 0 0 0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 CSH High ISLOCA None Low 1 1 0 0.00E+00 0.00E+00 0.OOE+00 0.00E+00 CSH Low Class 2 LSS N/A Assume Medium 13 0 -13 4.42E-10 4.42E-10 1.82E-11 1.82E-11 TOTAL 4.43E-10 4.77E-10 1.84E-11 2.54E-11 RHS High LOCA None Low 3 5 2 -1.70E-12 -1.70E-I2 -3.50E-13 -3.50E-13 RHS High PLOCA None Low 7 0 -7 3.50E-12 3.50E-12 3.50E-13 3.50E-13 RHS High ISLOCA None Low 4 1 -3 1.50E-11 1.50E-11 1.50E-11 1.50E-11 RHS High 21SLOCA None Low 0 1 1 -5.00E-13 -5.00E-13 -5.00E-14 -5.OOE-14 RHS Low Class 2 LSS N/A Assume Medium 72 0 -72 2.45E-09 2.45E-09 1.01E-10 1.01E-10 TOTAL 2.46E-09 2.46E-09 1.16E-10 1.16E-10 CSL High LOCA None Low 4 1 -3 2.55E-12 2.55E-12 5.25E-13 5.25E-13 CSL High PLOCA None Low 0 0 0 0.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 CSL High ISLOCA None Low 1 1 0 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 CSL Low Class 2 LSS N/A Assume Medium 9 0 -9 3.06E-10 3.06E-10 1.26E-11 1.26E-11 TOTAL 3.09E-10 3.09E-10 1.31 E-1 1 1.31 E-1 1 MSI High DTM - 1 / MSI - 1 None Low 0 5 5 -2.50E-12 -2.50E-12 -2.50E-13 -2.50E-13 TOTAL -2.50E-12 -2.50E-12 -2.50E-13 -2.50E-13 ICS High LOCA None Low 2 1 -1 8.50E-13 8.50E-13 1.75E-13 1.75E-13 ICS High ILOCA None Low 0 0 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 ICS High ILOCA - OC None Low 0 1 1 -1.70E-12 -1.70E-12 -7.OOE-14 -7.OOE-14 ICS Low Class 2 LSS N/A Assume Medium 10 0 -10 3.40E-10 3.40E-10 1.40E-11 1.40E-11 TOTAL 3.39E-10 3.39E-10 1.41 E-11 1.41 E-11 WCS High LOCA None(FAC) Low (High) 0 1 1 -8.50E-13 -8.50E-13 -1.75E-13 -1.75E-13 WCS High LOCA None Low -9 8 -1 8.50E-13 8.50E-13 1.75E-13 1.75E-13 WCS High ILOCA None Low 3 1 -2 1.00E-12 1.OOE-12 1.OOE-13 1.OOE-13 WCS High ILOCA - OC None Low 1 1 0 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 TOTAL 1.OOE-12 1.OOE.12 1.OOE-13 1.00E-13 to RBG-46922 Page 21 of 68 Table 3.4-1 (Cont'd)

Risk Impact Analysis Results System~() Safety Significance Break B

Location. 2 T DMs Failure Potential Rank (3) SXI(4)

Inspections RIS_B(5 ) Delta CDF Impact w/ POD w/o POD LERF Impact w/POD w/o POD DTM High LOCA None Low 0 3 3 -2.55E-12 -2.55E-12 -5.25E-13 -5.25E-13 DTM High ILOCA None Low 1 0 -1 5.OOE-13 5.OOE-13 5.OOE-14 5.OOE-14 DTM High MSD-3 None Low 0 3 3 z1.50E-12 -1.50E-12 -1.50E-13 -1.50E-13 DTM High DTM - 1 / MSI - 1 None Low 0 3 3 -1.50E-12 -1.50E-12 -1.50E-13 -1.50E-13 TOTAL -5.05E-12 -5.05E-12 -7.75E-13 -7.75E-13 GRAND GRTOTAL 3.97E-09 4.42E-09 2.05E-10 2.86E-10 1

Notes (1) Systems are described in Table 3.1.

(2) The "Class 2 LSS" break location designation in Table 3.4-1 is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).

(3) 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 (i.e., "Assume Medium").]

(4) 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.

(5) Inspection locations selected for RISB purposes that are in the plant's augmented inspection program for IGSCC are subject to the requirements provided below dependent upon other damage mechanisms identified. These requirements dictate how these inspection locations are accounted for in the risk impact analysis. -

(a) TACSC, TT, (IGSCC) and TT, (IGSCC) Damage Mechanism Combinations - these inspection locations are susceptible to thermal fatigue damage mechanisms in addition to IGSCC. In these cases, inspection locations selected for examination by both the IGSCC and RIS_B Programs should be included in both counts, but only those locations that were previously being credited in the Section XI Program and are now being credited in the RIS_B Program. The examination performed for IGSCC is judged adequate to have detected the other damage mechanisms subsequently identified by the RISB Program. For the RBS RIS_B application, one of these inspection locations was selected for examination per the plant's augmented inspection program for IGSCC and for RISB purposes due to the presence of other damage mechanisms. This inspection location was previously credited in the Section XI Program.

(a) None (IGSCC) Damage Mechanism - these inspection locations are susceptible to IGSCC only. In these cases, inspection locations selected for examination by both the IGSCC and RIS_B Programs should be included in both counts, but only those locations that were previously credited in the Section XI Program and are now being credited in the RISB Program. For the RBS RIS_B application, two of these inspection locations were selected for examination per the plant's augmented inspection program for IGSCC and are being credited for RISB purposes. These two inspection locations were previously credited in the Section XI Program.

to RBG-46922 Page 22 of 68 Table 5 Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 System(1 )

RPV Safety Significance High Low Break LOCA f DMs Failure Potential TASCS, TT, (IGSCC)

T 2 Rank" )

Medium (Medium)

Code Category B-F Weld Count 3

Section XI 3 0 Code Case N-716 Vol/Sur SurOnly RISB Other(3) 1(4) -

RPV LOCA TASCS,TT Medium B-J 1 1 0 1 -

Medium (Medium)

B-F 1 1 0 0 -

LOCA. TT, (IGSCC)

RPV B-J 1 1 0 0 -

B-F 18 18 0 1() -

RPV / LOCA None (IGSCC) Low (Medium)

B-J 4 4 0 1(5)

RPV V" LOCA None Low B-J 12 4 2 0 -

RDS V Class 2 LSS N/A Assume Medium C-F-2 48 4 0 0 -

RCS LOCA None Low B-J 114 0 0 12 -

FWS V" LOCA TASCS, TT, (FAC) Medium (High) B-J 25 15 0 5 -

FWS V LOCA TASCS,TT Medium B-J 28 16 0 0 -

FWS V LOCA TT Medium B-J 1 0 0 0 FWS V ILOCA - FW TASCS,TT Medium B-J 2 0 0 0 FWS V ILOCA - OC TASCS,TT Medium B-J 10 1 2 2 FWS " Class 2 LSS N/A Assume Medium C-F-2 23 4 0 0 MSS V LOCA None - Low B-J 111 8 12 10 MSS V ILOCA None Low B-J 4 1 0 2 MSS V ILOCA - OC None Low B-J 17 11 10 3 MSS V/ PLOCA None Low B-J 2 0 1 0 MSS V" DTM - I/MSI- 1 None Low B-J 8 0 8 0 MSS V/ - Class 2 LSS N/A Assume Medium C-F-2 16 2 0 0 to RBG-46922 Page 23 of 68 Table 5 (Cont'd)

Inspection Location Selection Comparison Between ASME Section Xl Code and Code Case N-716 System(1 ) Safety Significance Break Location Failure Potential Code Weld Section XI Code Case N-716 High Low DMs Rank(2 ) Category Count Vol/Sur Sur Only RIS_B Other(3)

SLS LOCA None Low B-J .17 0 13 4 -

SLS PLOCA None Low 'B-J 18 0 12 0 -

SLS ISLOCA None Low B-J 5 0 2 0 -

SLS V 21SLOCA None Low B-J 23 0 3 3 -

CSH V LOCA TT Medium B-J 4 3 0 1 -

CSH LOCA None Low B-J 1 1 0 0 -

CSH " PLOCA None Low B-J 10 0 0 0 -

CSH V ISLOCA None Low B-J 4 1 0 1 -

C-F-1 7 3 0 0 -

CSH Class 2 LSS N/A Assume Medium C-F-2 122 10 0 0 -

RHS V LOCA None Low B-J 13 3 0 5 -

RHS / PLOCA None Low B-J 37 7 0 0 -

RHS V ISLOCA None Low B-J 14 4 0 1 -

RHS V 21SLOCA None Low B-J 2 0 0 1 -

C-F-1 81 15 0 0 -

RHS " Class 2 LSS N/A Assume Medium C-F-2 785 57 0 0 -

CSL " LOCA None Low B-J 5 4 0 1 -

CSL " PLOCA None Low B-J 8 0 0 0 -

CSL " ISLOCA None Low B-J 5 1 0 1 -

C-F-1 5 4 0 0 -

CSL / Class 2 LSS N/A Assume Medium C-F-2 73 5 0 0 -

MSI / DTM- 1/MSI- 1 None Low B-J 43 0 26 5 -

to RBG-46922 Page 24 of 68 Table 5 (Cont'd)

Inspection Location Selection Comparison Between ASME Section Xl Code and Code Case N-716 System(l) Safety Significance Break Location Failure Potential Code Weld Section XI Code Case N-716 High Low DMs Rank(2 ) Category Count Vol/Sur Sur Only RISB Other(3 )

ICS LOCA None Low B-J 12 2 0 1 -

ICS ILOCA None Low B-J 2 0 0 0 -

ICS ILOCA - OC None Low B-J 5 0 0 1 -

C-F-1 6 0 0 0 -

ICS Class 2 LSS N/A Assume Medium C-F-2 159 10 0 0 -

WCS VY LOCA None (FAC) Low (High) B-J 4 0 1 1 -

WCS V LOCA None Low B-J 81 9 15 8 -

B-J 4 3 0 1 -

WCS V ILOCA None Low C-F-2 9 0 0 0 -

Class 3 1 0 0 0 -

B-J 2 0 0 1 -

WCS ILOCA- OC None Low C-F-2 4 1 0 0 -

Class 3 3 0 0 0 -

DTM VI LOCA None Low B-J 27 0 15 3 -

DTM " ILOCA None Low B-J 3 1 2 0 -

DTM " MSD-3 None Low B-J 9 0 6 3 -

DTM " DTM - 1/MSI- 1 None Low B-J 43 0 6 3 -

Notes (1) Systems are described in Table 3.1.

(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 (i.e., "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. RBS selected a 10% sampling without relying on IGSCC Program locations beyond those selected for RIS+B purposes either due to the presence of other damage mechanisms, or where no other damage mechanism is present. The "Other" column has been retained in this table solely for uniformity purposes with other RISB application template submittals.

(4) This piping weld selected for examination per plant augmented IGSCC inspection program (Category C) and RISB purposes due to' presence of other damage mechanisms.

(5) Two piping welds (Code Category B-F and B-J) selected for examination per the plant augmented IGSCC inspection program (Category C) and RIS_B purposes.

to RBG-46922 Page 25 of 68 APPENDIX I River Bend Station PRA Model Capability for Use in Risk-Informed Inservice Inspection Applications Introduction The River Bend Station Probabilistic Risk Assessment (PRA) was initially developed in response to NRC Generic Letter (GL) 88-20, Individual Plant Examinations (IPE's). The, IPE was submitted to the NRC via letter RBG-38077 dated 15 January 1993. The RBS IPE consisted of the Level 1 PSA, including addressing internal flooding, and the back-end analysis (Level 2) consistent with the requirements of GL 88-20, Individual Plant Examination for Severe Accident Vulnerabilities. The NRC provided a Staff Evaluation in a letter dated October 17, 1996, which approved the RBS IPE. The Staff Evaluation concluded that the RBS IPE met the intent of GL88-20, that is, the RBS IPE process was capable of identifying the most likely severe accidents and severe accident vulnerabilities for RBS.

Several PRA model updates have been completed on the RBS PRA since the IPE was submitted. These were done to maintain the PRA model reasonably consistent with the as-built as-operated plant. The scope of the updates was based on review of results, plant input to the model, updated plant and component failure and initiating event data, modifications to plant design, changes to plant procedures, as well as model enhancements. As part of major updates, an internal review of PRA model results is performed utilizing an expert panel composed of experienced personnel from various plant organizations, including Operations and Engineering.

The RBS PRA has been used as a basis for risk-informed submittals to the NRC, including_

Amendment 125 to the RBS license, approved by a Sept. 25, 2002, NRC letter. While the NRC did not review the RBS PRA, the staff had asked River Bend to perform various calculations, the results of which caused the NRC staff to agree with the overall assessment of the previous 1998 BWROG peer review that the RBS PRA was suitable for supporting risk-informed applications.

The NRC found that the RBS PRA was adequate to support a RG 1.177 risk assessment.

The RBS PRA is currently at Revision 4.a. This minor revision was approved in March 2008 and implemented a model for cooling of the Control Building switchgear rooms. Core Damage Frequency (CDF) is predicted to be 3.55E-06 per year with a truncation limit of 1 E-1 1/year.

The previous full revision of the RBS PRA was Revision 4. This revision was approved in September 2005. The Rev.4 CDF was calculated to be 1.94E-06/year with a truncation limit of 1E-1 0/year. The Large Early Release Frequency (LERF) for this model was calculated to be 2.53E-08/year. Model changes incorporated in the March 2008 update result in minimal impact (approximately 10%) upon LERF results.

As discussed below, the RBS PRA is more than adequate for this RI-ISI application. The PRA model used for this application has been evaluated against RG 1.200 Revision 1 and all gaps with respect to RG 1.200 have been evaluated. Most of the gaps are documentation issues.

The few remaining gaps which could have been potentially applicable to use of the model for RI-ISI have been successfully addressed, as documented herein. It is concluded that the RBS PRA model fully supports the needs of this RI-ISI submittal, as the internal flooding calculation CDF and LERF results for each scenario are well below the risk thresholds for ASME Code Case N-716.

to RBG-46922 Page 26 of 68 PRA Maintenance and Update Entergy employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Entergy nuclear plants. This approach includes both a proce~duralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews.

The Entergy risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plant. Thisprocess is defined in the Entergy Risk Management program. The overall Entergy Risk Management program, through procedure EN-DC-151, "PSA Maintenance and Update", defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:

Design changes and procedure changes are reviewed for their impact on the PRA model.

New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

Maintenance unavailabilities are captured, and their impact on CDF is updated as part of the model revision process.

Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four to five years.

In addition to these activities, Entergy risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes the following:

Documentation of the PRA model, PRA products, and bases documents.

(Procedures EN-DC-126 and EN-DC-1 51.)

Guidelines for updating the full power, internal events PRA models for Entergy nuclear generation sites. (Procedures EN-DC-126 and EN-DC-151i.)

Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (1 OCFR50.65 (a)(4)). (River Bend.Station Procedure ADM-0096)

Issues requiring action are entered into the Model Change Request (MCR) database which is controlled under EN-DC-151. These issues are prioritized in accordance with their significance for implementation into future PRA updates. Significant issues that are a result of errors are entered into the Entergy corrective action program under EN-LI-1 02.

to RBG-46922 Page 27 of 68 In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately four to five year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant. Entergy performs regularly scheduled updates to the RBS PRA model. The next RBS PRA model update is scheduled for 2009-2010 and is expected to be approved in 2010.

PRA Self Assessment and Peer Review RG 1.178 [10] specifies that a description of industry reviews performed on the PRA be included. Therefore, the independent PRA Peer Review and the ASME PRA Standard self-assessment review are included here along with the resolution of the review comments.

Several assessments of technical capability have been made, and continue to be planned for RBS PRA model. These assessments are as follows and further discussed in the paragraphs below.

An independent PRA peer review was conducted under the auspices of the BWR Owners' Group in November 1998, following the industry PRA Peer Review process [1] which was the basis for the industry PRA Peer Review process NEI 00-02. The predecessor to the ASME PRA Standard Peer Review process was NEI-00-02 which identified the critical PRA elements and their attributes necessary for a quality PRA. This peer review included an assessment of the PRA model maintenance and update process.

• During 2005 and 2006, the RBS PRA model results were evaluated in the BWR Owners' Group PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.

In 2008, a self-assessment analysis [14] was performed against the available version of the ASME PRA Standard [2] and Regulatory Guide 1.200, Rev. 1 [9].

As part of the PRA model update scheduled for 2009-2010 and expected to be approved in 2010, the self-assessment analysis will be updated to reflect pertinent changes to both the PRA Standard and Regulatory Guide 1.200.

Internal Flooding Model The RBS Internal Flooding Analysis (IFA) was significantly upgraded to meet the requirements of RG 1.200 in 2009. This analysis was used in the subject RI-ISI evaluation to determine the High Safety-Significant (HSS) scope and as an input to Low Safety-Significant (LSS) scope Conditional Core Damage Probability (CCDP) values used in the risk impact assessment. This analysis is a substantial improvement over the previous IPE version of the Internal Flooding PRA (IFPRA). As an example, the IPE IFA conservatively used a 1 E-7 screening value, and no scenarios resulted in CDF higher than the screening value. The current IFA has approximately 500 quantified scenarios;, with CDF's ranging from about 2E-7 to less than 1E-12 range. Many of the scenarios have a CDF lower than the quantification truncation value used (1E-12).

Further, due to the conservative simplifications required to analyze the large number of scenarios for an Internal Flooding PRA, these results are considered to be more conservative in inherent nature compared to the base Internal Events PRA. Other improvements including accounting for all liquid systems (e.g., Fire Protection) as flood sources, use of improved pipe rupture frequency data, and accounting for the frequency of breaks of up to a full guillotine break in the affected piping.

to RBG-46922 Page 28 of 68 PRA Peer Review An independent assessment of the RBS PRA, using the Self-Assessment Process developed as part of the Boiling Water Reactor Owners' Group PRA Peer Review Certification Program, was completed in 1998. Certification was performed by a team of independent PRA experts from U.S. nuclear utility PRA groups and PRA consulting organizations. This intensive peer review involved about two man-months of engineering effort by the review team and provided a comprehensive assessment of the strengths and limitations of each element of the PRA. The peer review concluded that the PRA was suitable for supporting risk-informed applications, such as Technical Specification changes, provided some enhancements were made, including four of High significance and numerous of less significance. Items evaluated through a peer review which were considered to not meet technical element requirements are documented with F&O's.

Each F&O is provided with a level of significance (A: Extremely Important; B: Important; C:

Desirable; D: Minor). The high significance items and all but a few of the lower significance items had been adequately addressed by the time the NRC approved an extension to the Allowed Outage Time (AOT) for RBS Emergency Diesel Generators via Amendment 125 to the RBS License (NRC letter dated Sept. 25, 2002).

In 2006, a summary of the disposition of Industry PRA Peer Review facts and observations (F&Os) arising from the BWROG PRA peer review for the RBS PRA model was documented as part of the statement of PRA capability for MSPI in the RBS MSPI Basis Document [4]. As noted in the RBS MSPI Basis Document, all of the "A" level F&Os identified in the PRA Peer Review were addressed and resolved in the RBS update model (Revision 4) approved for use in 2005. In addition, all of the "B" level F&Os were resolved within the model except for two.

Entergy subsequently demonstrated that these two F&Os were insignificant for inclusion in the

,MSPI evaluation. Also noted in the MSPI .Basis Document was the fact that, after allowing for plant-specific features, there are no MSPI cross-comparison outliers for RBS. Since these

F&O's are insignificant with respect to MPSI, they also do not impact the ability to use the RBS PRA model in support of RIISI applications. Table 3 addresses the two remaining "B" level F&O items.

Self-Assessment A 2008 Self-Assessment analysis for the RBS PRA model approved in 2005, including a March 2008 minor revision, was completed in December 2008. [14] This Self-Assessment analysis was performed against ASME PRA Standard [13] as endorsed by Regulatory Guide 1.200 Revision 1 [9]. This self-assessment analysis identified a list of 72 supporting requirements from the Standard which did not meet the Standard.

Table 1 provides the results of the River Bend Self-Assessment and identifies those ASME PRA Supporting Requirements that could require a sensitivity study or other disposition to more fully support the RI-ISI analysis. Table 2 provides a disposition of these identified gaps, including discussion of applicable sensitivity calculations.

The River Bend Station Internal Flooding PRA has recently been updated and meets ASME PRAStandard requirements for Internal Flooding (IF), including documenting compliance with all ASME standard supporting requirements. Entergy has also conducted an internal review of its processes* to ensure they meet the ASME Standard supporting requirements for Maintenance and Update, (MU).

to RBG-46922 Page 29 of 68 Plant modifications which have not yet been incorporated into the RBS PRA and other potential issues identified since the model was revised have been reviewed. It is concluded that none of these items would significantly impact the ability of the RBS PRA to support a RI-ISI application.

Self-Assessment Interpretation PRAs can be used in applications despite not meeting all of the Supporting Requirements (SRs) of the ASME PRA Standard. This is well recognized by the NRC and is explicitly stated in the ASME PRA Standard and RG 1.174. RG 1.174 states the following in Section 2.2.6:

There are, however, some applicationsthat, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

An example is risk-informed in-service inspection (RI-ISI). In this application, risk significance was used as one criterion for selecting pipe segments to be periodicallyexamined for cracking. During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary.

Therefore, the staff review of plant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability.

Therefore, a RI-ISI PRA application requires no more than Capability Category I.

It is also acknowledged that for PRAs with SRs ranked as "Not Met", the PRA may be used for PRA applications but may require additional justification and support to allow their use.

Finally, it is judged that no PRA has Capability Category III for all of its SRs, nor is this currently expected as part of the NRC PRA Quality Program:

A review is performed of these Supporting Requirements (SRs) in Table 1 assessed as "Not Met" based on the self-assessment. The evaluation, disposition, and justification for the 72 "Not Met" supporting requirements is included in Table 1. The vast majority of the "Not Met" supporting requirements are documentation deficiencies rather than technical issues with the model itself. The importance of the Supporting Requirements to the RI-ISI application considers a spectrum of possible outcomes. For these "Not Met" SRs, they are dispositioned as follows:

• PA: Potentially applicable. Sensitivity Case may be required.

• NA: Not applicable. Areas of model that are not used in the RI-ISI evaluation

• NS: Not significant for the RI-ISI PRA application. Associated with areas that have no effect on the RI-ISI process or the risk significance determination, e.g.:

Areas that are clearly not significant to the quantification of the risk for welds are:

Strictly documentation issues Parametric uncertainty analyses Modeling uncertainties not part of the LOCA scenarios

• NO: None. Meets at least Capability Category I

• SI: Specific Issue. A specific technical issue could influence the risk assessment as it affects one or more systems.

to RBG-46922 Page 30 of 68 For those SRs that are "Not Met" and have a potential impact on the RI-ISI evaluation, a sensitivity case is defined that would demonstrate the capability of the PRA to appropriately characterize the PRA for use in RI-ISI to meet the ASME PRA Standard and RG 1.200.

Table 1 summarizes the disposition of the "Not Met" Supporting Requirements (SRs). As can be seen, approximately 66% of the "Not Met" SRs are related to documentation or modeling uncertainty assessments. These are expected to have no significant impact on the RI-ISI evaluation. Six SRs are identified for additional sensitivity cases. These six SRs subsume an additional two SRs. Table 2 provides a disposition for these gaps which demonstrates that the River Bend PRA provides a basis of sufficient quality to support the RI-ISI application. This includes documentation of sensitivity cases performed in response to the RG1.200 PRA self-assessment.

General Conclusions Regarding PRA Quality for RI-ISI:

The RBS PRA maintenance and update processes and technical capability evaluations described above provide a robust basis for concluding that the PRA is suitable for use in RI-ISI risk-informed licensing actions. In the risk-informed inservice inspection program at RBS, the EPRI Risk Informed ISI methodology (Reference 7) is used to define alternative inservice inspection requirements. Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking, element selection, and risk differential evaluation steps.

The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental components of the EPRI methodology. First, PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection as illustrated below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results. Further, the LSS classifications were conservatively binned as High Risk. None of the Medium Risk break locations challenged the High classification (Highest was 3.4E-04 for CCDP and 1.4E-05 for CLERP).

The risk importance of a weld is therefore not tied directly to a specific PRA result. Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology. Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population. Secondly, the impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix.

Thirdly, the EPRI RI-ISI methodology uses an absolute risk ranking approach. As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspection population rather than masking other important components. That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population. These three facets of the methodology reduce the importance and influence of PRA on the final list of candidate welds.

to RBG-46922 Page 31 of 68 Conclusion Regarding PRA Capability for RI-ISI:

The RBS PRA models are suitable for use in the RI-ISI application. This conclusion is based on,

• The PRA maintenance and update process in place,

• The PRA technical capability evaluations that have been performed and are being planned, and The RI-ISI process considerations, as noted above, that demonstrate the relatively limited sensitivity of the EPRI RI-ISI process to PRA attribute capability beyond ASME PRA Standard Capability Category I.

As the PRA analysis continues to be improved during the 10-year interval, these results will be reviewed to determine which, if any, would merit RI-ISI specific sensitivity studies.

References

1. Boiling Water Reactors Owners' Group, BWROG PSA PeerReview Certification Implementation Guidelines, Revision 3, January 1997.

2. American Society of Mechanical Engineers, Standard for ProbabilisticRisk Assessment for NuclearPower PlantApplications', ASME RA-S-2002, New York, New York, April 2002.

3. U.S. Nuclear Regulatory Commission, An Approach for Determiningthe Technical Adequacy of ProbabilisticRisk Assessment Results for Risk-Informed Activities, Draft Regulatory Guide DG-1 122, November 2002.

4. River Bend MSPI Basis Document, RBS-SA-06-00001. Rev. 2, dated April 5, 2007.

5. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sb-2005, New York, New York, December 2005.

6. U.S. Nuclear Regulatory Commission Memorandum to Michael T. Lesar from Farouk Eltawila, "Notice of Clarification to Revision 1 of Regulatory Guide 1.200," for publication as a Federal Register Notice, July 27, 2007.

7. Revised Risk-Informed Inservice Inspection Evaluation Procedure,EPRI TR-1 12657, Revision B-A, December 1999.

8. U.S. Nuclear Regulatory Commission, An Approach for Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 1, November 2002.

9. NRC Regulatory Guide, RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Rev. 1, January 2007.

10. An Approach for Plant-Specific Risk-Informed Decision making for In service Inspection of Piping, Regulatory Guide 1.178, Rev. 1, US NRC, September 2003.

11. Alternative Piping Classification and Examination Requirements (Section XI, Division 1),

Case N-716, Cases of ASME Boiler and Pressure Vessel Code, Approval Date: April 19, 2006.

to RBG-46922 Page 32 of 68

12. Standard Review Plan for the Review of Risk-Informed Inservice Inspection of Piping, Standard Review Plan Office of Nuclear Reactor Regulation, NUREG-0800, September 2003.

13. ASME RA-Sb-2005, Addenda to ASME RA-S-2002, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, December 2005.

14. River Bend PRA Self-Assessment, Report C247080005-8620, Rev. 2, dated February 2, 2009.

15. Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs, EPRI, 1018427, December 2008.

16. PRA Quality for RI-ISI PRA Application, Report C247080010-8506 dated February 4, 2009.

Attachment 1 to RBG-46922 Page 33 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs Importance to RI-ISI Gap #1 Initiators appropriately categorized and plant specific features IE-A2 accounted for. " NS: Not Significant The implementation of the ISLOCA evaluations in PRA-RB Category I pipes are already placed in the High Safety Significant 002S08 has assumed that any 4 valves in a line qualify to allow Category. Therefore, this SR is judged not to result in changing screening a line from consideration. This is judged to be the HSS categorization of the interfacing lines inside the 2nd isolation valve.

inconsistent with NSAC-154 and typical PRA practice. The isolation valves that are to be counted must:

(1) be able to close against the differential pressure

" PA: Potentially Applicable or For lines outside the 2 nd isolation valve in high pressure lines (e.g.,

(2) must be closed as their normal position main steam lines), use bounding estimates of the valve interface The lines screened are the 3 LPCI and 1 LPCS injection lines. rupture isolation capability, pipe rupture frequency, and the CCDP(ý) to assess whether the pipe segments need to be added Low pressure rated pipe in the LPCI, SDC, and LPCS systems (See IE-C4 for breaks outside containment in high energy lines).

has been hydrostatically tested at rated RPV pressure.

Discussion with Entergy [16] indicated that the low pressure pipe for LPCI and LPCI has been hydrostatically tested at normal operating pressures of the RPV (>1000 psig).

Therefore, there is high confidence that the pipe is capable of withstanding any potential high pressure condition that could result from a failure of the high pressure to low pressure interface valves. Based on this plant unique resolution to the ISLOCA question, no additional sensitivity cases are needed.

However, this information should be documented in the PRA to support the ISLOCA evaluation..

Breaks outside containment in high energy lines beyond the 2 nd isolation valve (Main Steam, FW, HPCS, RWCU, RCIC) are also in need of evaluation to ensure these are properly accounted for. (See IE-C4) ____________ w (1) CCDP = Conditional Core Damage Probability

Attachment 1 to RBG-46922 Page 34 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap J Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI Gap #1 Code Case N-716 allows the use of bounding estimates for both -

(cont'd) the pipe rupture frequency and the CCDP. See Section 5 of Code Case N-716. With this allowance, the pipe segments that lie outside the 2nd isolation valve for ECCS, RWCU, MS, and FW can be conservatively evaluated to determine if these pipe segments have a sufficiently low risk contribution to be placed in the low safety significant category.

NS: Not Significant

• Reference Leg leak-down is an initiating event that can Reference leg leakdown should be explicitly evaluated as part of a compromise multiple systems. This initiator should be identified PRA update but this is not required for RI-ISI application because for disposition. A loss of Ref. Leg is not assessed. (See SLI- it has a negligible influence on pipe break frequency and the 8211 [1], SL18218 [2], SL18221 [3] for typical approaches used overall-risk profile.

in BWR PRAs).

Gap #2 Challenges to safe shutdown from power conditions occurring at IE-A5 NS: Not Significant power levels other than 100% power are included in the Initiating None of the events eliminated from RBS consideration would Events Assessment. Examples of non-applicable events which are significantly influence the assessment of pipe failure frequencies.

not included: refueling events, external events. As such, the RI-ISI can be performed effectively without the resolution of this SR to a higher Capability Category.

See plant specific initiating event data assessment.

However, the exclusion of the loss of transformer event at 10%

power from the plant specific assessment is contrary to IE-A5 requirement to incorporate such events.

Exclusion of LERs should be reconsidered in light of the ASME SR IE-A5 to consider events if they could have occurred at power specifically April 11, 2003 outage #FO 03-02.

The methodology section does not discuss how LP/SD events are addressed in the analysis.

Gap #3 Refer to SR IE-C4 regarding screening of excessive LOCA and IE-B4 See IE-C4.

assessment of Breaks Outside Containment (BOC) initiators.

Attachment 1 to RBG-46922 Page 35 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap [ Description of Gap to Capability Category II Applicable SRs ] Importance to RI-ISI Gap #4 The resulting IE frequencies are in units of critical years; the IE-C3 NO: None posterior critical year values were not converted to reactor This change would decrease the initiating event frequency and calendar year using the predicted plant availability as required by resulting CDF and LERF. As such, the RI-ISI can be performed SR IE-C3. effectively without the resolution of this SR to a higher Capability In the next PRA Update, update the Initiating Event Notebook Category.

(River Bend PSA-001) so that it addresses the expected future plant availability and how that relates to the availability used in the EPRI [15] has reviewed the ASME PRA Standard Supporting predictive IE frequency calculations. Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISl[application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #5 Initiating events within the screening criteria are retained. IE-C4 See IE-A2.

Initiating events are in general appropriately subsumed into the EPRI [15] has reviewed the ASME PRA Standard Supporting resultingin initiating criteria the ASME event categories.

Standard are notTherefore, the screening explicitly used R irements for r theirtheapplicability a SME in in ensuring e nsu r the ing to eliminate Requirements technical quality of initeriaintor from considerao arRI-ISI initiators from consideration.-- risk-informed decisions. Based on the EPRI evaluation [15],

this Supporting Requirement does not need to be met in order to The implementation of the ISLOCA evaluations in PRA-RB adequately support the EPRI RI-ISI application methodology. This 002S08 has assumed that any 4 valves in a line qualify to allow evaluation is also judged to apply to the methodology used in the screening a line from consideration. This is judged to be Code Case N-716 RI-ISI.

inconsistent with NSAC-154 and typical PRA practice. The isolation valves that are to be counted must:

(1) be able to close against the differential pressure or (2) must be closed as their normal position The lines eliminated incorrectly are the 3 LPCI and 1 LPCS injection lines.

Attachment 1 to RBG-46922 Page 36 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Gap #5

• Description of Gap to Capability Category II For SDC Suction Line I Applicable SRs [ Importance to RI-ISI

• NS: Not Significant (cont'd) The probability of line failure given the interface fails should - This would result in decreasing CDF. As such, the RI-ISI can also be assessed probabilistically instead of setting to 1.0. be performed effectively without the resolution of this SR to a higher Capability Category.

• Vessel rupture has been excluded from consideration based - Vessel rupture has negligible influence on the CDF solely on a frequency estimate. This is not consistent with SR IE-B4 and IE-C4 requirements. - Category I pipes are already placed in the High Safety

• Breaks Outside Containment (BOC) in high energy lines are Significant Category. Therefore, this SR is judged not to result not evaluated for their impact and frequency. in changing the HSS categorization of the interfacing lines inside the 2nd isolation valve except as noted in IE-A2.

Gap #6 List of generic priors is consistent with "NRC Issue" expectation. IE-C10 NS: Not Significant However, the "NRC Resolution" for this SR also requires that Documentation related item to explain differences.

differences be explained.

Gap #7 "Loss of intake" initiators and associated plant-specific issues not IE-C1 1

• NS: Not Significant discussed in the IE analysis. This is a documentation issue. The model is not being changed to RBS will want to document CR-RBS-2007-4447 loss of makeup address this item.

event, for which there was a significant and quick down power, but for which the plant did not scram or go off-line.

Gap #8 All of the ISLOCA features are not explicitly addressed in the IE-C12 See IE-A2.

derivation of the ISLOCA frequency. Items (b) - (e) are not included in the ISLOCA analysis.

Gap #9

• Modeling uncertainties are not discussed or identified. IE-D3 NS: Not Significant The RBS PRA includes a substantial number of sensitivity

• Peer review expectations and an edict from the BWROG to calculations that demonstrate the range of uncertainties declare these SRs as Not Met will in general lead to "Not Met" associated with specific assumptions and modeling uncertainties.

categorization by Peer Review Teams. These sensitivity studies are consistent with the expected NUREG-1855 approach to sensitivity analyses for evaluation of modeling uncertainties.

To be determined once the new NRC/EPRI guidance is available (e.g., NUREG-1855). However, the EPRI RI-ISI process is defined such that model uncertainties will not unduly influence results.

Attachment 1 to RBG-46922 Page 37 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs I Importance to RI-ISI Gap #9 EPRI [15] has reviewed the ASME PRA Standard Supporting (cont'd) Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #10 Insufficient information included to ascertain this, e.g., effect of AS-B5a INS: Not Significant RHR in SPC during power operation. Subsequent Entergy input [16] indicated that:

Many different alignments are considered in the PRA. The particular alignment for SPC initially running when LPCI is actuated is not normally an issue for BWR-6 plants. At best, this is a Specific Issue rather than a Potentially Significant issue.

Informal documentation from 1998 indicates the water hammer issue was not considered an issue at RBS due to'the difference in elevations between the high point in the line and the suppression pool lower limit was not sufficient to result in voiding. Thus, no significant water hammer would occur.

This becomes a documentation issue that needs to be addressed under separate SRs.

N

Attachment 1 to RBG-46922 Page 38 of 68 TABLE I ,

STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap I Description of Gap to Capability Category II 1 Applicable SRs I Importance to RI-ISI Gap #11 PRA-RB-01 -002S01 provides some information. AS-Cl 0 NS: Not Significant The RBS PRA includes a substantial number of sensitivity Event Trees that appear to be missing include (p. 58 of 78):

calculations that demonstrate the range of uncertainties

• ISLOCA associated with specific assumptions and modeling uncertainties.

• ATWS These sensitivity studies are consistent with the expected

• Internal Flood NUREG-1855 approach to sensitivity analyses for evaluation of modeling uncertainties.

• The event trees are not presented in the Accident Sequence Calc PRA-RB-01-002S01 The RBS approach is to be determined once the new NRC/EPRI

• The functional fault trees that relate the Event Tree Nodes guidance is available (e.g., NUREG-1855). However, the EPRI to the systems supporting the functions are not presented. RI-ISI process is defined such that model uncertainties will not

• The individual sequence descriptions are presented but not unduly influence results.

the event tree.

Event Trees reviewed as part of Internal Flooding analysis.

There appears to be significant gaps in the documentation Internal Events PRA event tree confirmed to be appropriate for regarding how the model is assembled. The gaps are primarily in Internal Flooding.

the following areas:

" The connection between systems (which have relatively limited documentation) and the function used in the event tree nodes. Examples include the following:

- LOSP in the transient event tree

- X1: RPV depressurization for different sequences (e.g.,

initiators)

> Transients, Small LOCA, Med. LOCA

- V2, V3 for transients or Large LOCA

• The event trees do not specify the branch fault trees used in the quantification; e.g.,

- RCIC for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

- RCIC for4 hours

- Other

Attachment 1 to RBG-46922 Page 39 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II [ Applicable SRs - Importance to RI-ISI Gap #11

• The treatment of sequence transfers is not clear. Are the

• NS: Not Significant (cont'd) successes carried forward with the transfers; if so, how? A sensitivity case to extend the truncation level to that used for

• Transfer 4 for transient p. 4 does not appear to be other sequences is desirable but not required for RI-ISI.

connected. NA: Not Applicable

" The small LOCA Seq. 42 appears to be truncated at 3E- - Pipe inside containment (Class I) is already classified as HSS.

8/yr. This would appear to be inappropriate. This difference has no material effect on the pipe classification.

" For small LOCA, Seq. 9 & 10 appear to allow RCIC success - Restructure the SLOCA event tree to require depressurization for the 24 hr. mission time. and LP injection for SLOCA with RCIC and SORV with RCIC.

This would require crew alignment of RHR within a short time (2-4 hrs) to ensure HCTL is not exceeded.

0 In addition, there could be an SORV. RCIC is not a success for a 24 hr. mission with an SORV.

Gap #12 See Uncertainty Calculation (PRA-RB-01-002S13). AS-C3 NS: Not Significant The RG 1.200 endorsement of the ASME PRA Standard has The RBS PRA includes a substantial number of sensitivity included a requirement to document the assumptions and sources calculations that demonstrate the range of uncertainties of uncertainty associated with each PRA element. The NRC and associated with specific assumptions and modeling uncertainties.

the industry are working together to clarify what this means and to These sensitivity studies are consistent with the expected develop a structure that will satisfy these SRs. As of the NUREG-1855 approach to sensitivity analyses for evaluation of performance of this self-assessment, this cooperative effort has modeling uncertainties.

not been completed. The RBS approach is to be determined once the new NRC/EPRI Peer Review expectations and an edict from the BWROG to guidance is available (e.g., NUREG-1855). However, the EPRI declare these SRs as Not Met will in general lead to "Not Met" RI-ISI process is defined such that model uncertainties will not categorization by Peer Review Teams. - unduly influence results.

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Attachment 1 to RBG-46922 Page 40 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs I Importance to RI-ISI Gap #13 Not defined. SC-A2

• NS: Not Significant The parameter basis for determining success or failure of a This is a documentation issue. The model is not being changed to sequence is not defined address this item.

• Core Damage - Core temperature

• RPV - Pressure Capacity

• Containment - Pressure Capacity

- Hydrodynamic loading under ATWS conditions Gap #14 Mission times are discussed in Accident Sequence Calculation SC-A5 See SC-Al.

PRA-RB-01-002S01.

The mission times for failure to run calculations are assessed at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less if specifically justified.

Extending the FTR mission time beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for loss of DHR sequences is considered to be an unnecessary complication and does not affect PRA insights nor does it significantly affect its quantitative evaluation.

Attachment 1 to RBG-46922 Page 41 of 68 TABLE I STATUS OF."NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap I Description of Gap to Capability Category II ] Applicable SRs I Importance to RI-ISI Gap #14 The evaluation of safe stable states in a PSA has generally (cont'd) involved the assessment of equipment operation and operator actions over an extended period of time. This extended period of time is nominally taken to be sufficiently long such that offsite resources can be brought to bear to mitigate or further prevent accident progression. The considerations that have dominated the choice of the mission time are as follows:

" Equipment failure rates (failures/hour) are judged to be too conservative for times greater than a few hours of operation.

• For times greater than a few hours, the ability to repair and recover equipment can compete with the failure rate such that there can be considered to be a steady state equilibrium condition reached.

" For times greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the TSC and EOF would be manned, and additional expertise could be available by phone or transported to these facilities.

• For times greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, it is considered highly likely that offsite resources (e.g., equipment, power, vehicles) would be available as back-ups to primary methods of prevention and mitigation.

" From a risk perspective, actual data from natural and man-caused disasters have indicated that public evacuations can be effectively carried out in time frames of less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Therefore, prevention of accidents through 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of mission time have the largest potential for early health effects risk reduction.

Attachment 1 to RBG-46922 Page 42 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap [ Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI Gap #14

• Finally, beyond time frames of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, "ad hoc" (cont'd) procedures can be written and reviewed to perform alignments and equipment usage that are not part of current plant practices or training. Such ad hoc procedures and equipment usage can cover such a wide spectrum of possibilities that it is judged not useful to develop all possible contingencies at this time.

Based on the above considerations, it has been considered in past PSAs that it is to appropriate to use an equipment mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This consideration dictates the use of equipment "run" failure rates (per hour) coupled with a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time to calculate the "run" failure probability of equipment. This calculated "run" failure probability is then treated conservatively by applying this "run" failure probability as a failure that is postulated at time zero.

Attachment 1 to RBG-46922 Page 43 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap I Description of Gap to Capability Category II 1 Applicable SRs I Importance to RI-ISI Gap #15 The modeling has been performed consistent with the as-built, as- SC-A6

• NS: Not Significant operated plant as of the PRA modeling freeze date with the

- Credit for use of Service Water as a vessel injection source is possible exceptions: part of BWR EOP's. Grand Gulf justification is that the Service

" There is no basis presented for SSW as a successful Water flow is comfortably greater than LPCI injection flow. This injection source except the assertion that the EOP-0001 is also the case for River Bend, where Standby Service Water specifies its use. This is not an adequate basis. pumps are rated at 7690 gpm per SAR Table 9.2-15 whereas LPCI injection flow credited in accident analyses is 4470 gpm Based on discussions with Entergy [16], the SW system per SAR Appendix 15.b. Per Process Flow Diagrams has a discharge head of approximately 80 psig. This (0221.434-000-015, -016, -017), the SWP system is capable of head should be more than adequate to provide supplying at least 5800 GPM to the input of the RHR HX, significant flows to the RPV via the cross tie connection including flow through the tube side of both heat exchangers when the RPV is fully depressurized. Therefore, the and return flow to the Standby Cooling Towers, which is primary gap is associated with providing these facts in considered a higher resistance flow path than injection to the the PRA documentation (e.g:, success criteria notebook reactor vessel. Per TSG-001 (Severe Accident Procedure or SW system notebook). Technical Support Guidelines), each Standby Service Water

" The RCIC back pressure trip is taken at 25 psig and is pump is capable of providing 9600 gpm of flow (best estimate calculations; on same basis, LPCI can provide 5650 gpm).

then related to the pressure in containment at 25 psig.

There is generally a lower containment pressure In addition, based on discussions with Entergy [16], the SW associated with the trip. This is usually approximately 6 psig less or 19 psig as measured in containment. Based system has a discharge head of approximately 80 psig. This head should be more than adequate to provide significant flows on discussions with Entergy [16] the RCIC operation for to the RPV via the cross tie connection when the RPV is fully a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time requires RHR success in depressurized. Therefore, the primary gap is associated with suppression pool cooling. For SBO, the operation of RCIC is limited by a number of factors: providing these facts in the PRA documentation (e.g., success criteria notebook or SW system notebook). As such, this "gap"

- CST inventory is judged to be not significant for the assessment of the RI-ISI

- Suppression pool temperature applications.

- Battery capacity - RCIC success for extended SBO events has properly considered the availability of RHR and the potential for a Because of the potential limitations, RCIC operation for 6 shorter RCIC mission time.

hours is credited. RCIC is not credited beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the SBO conditions.

The RCIC high turbine back pressure trip (even for cases of recirc seal LOCA) is expected to occur beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and therefore, is not limiting.

Attachment 1 to RBG-46922 Page 44 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap I Description of Gap to Capability Category II t Applicable SRs I Importance to RI-ISI Gap #15 Success Criteria Tables in Appendix C of the Accident Sequence (cont'd) Calculation (PRA-RB-01 -002S01) appear to have the following additional needs:

• Specific deterministic calculations (e.g., MAAP) to " NS: Not Significant demonstrate success of systems This is a documentation issue. The model is not being changed to address this item.

" Additional clarifications to identify the mission time for

• NS: Not Significant success of systems such as RCIC (e.g., see p. 6 of 78 of This is a documentation issue. The model is not being changed to PRA-RB-01-002S01) as an example of a clarification on address this item.

RCIC mission time.

• The assumption that venting and 1 fan cooler is

• NS: Not Significant adequate for containment pressure control is not based This is a documentation issue. The model is not being changed to on reaching a safe stable state. It is said to be "more address this item.

stable" than the unstable case. However, the ability of containment coolers or venting to mitigate containment pressure rise is documented in Section 3.1 of PRA-RB-001-002S01. Given the significant effect each has in lengthening the time to containment failure, it would appear to be a reasonable engineering judgment that the two combined would result in preventing containment failure.

• The failure of HPCS in large and medium LOCAs if SPC

• NS: Not Significant is unavailable does not appear to be explained in a This is a documentation issue. The model is not being changed to clearly unambiguous fashion. It is not reflected in the address this item.

success criteria table.

" RCIC failure during loss of DHR may occur due to:

• NS: Not Significant (1) High turbine exhaust back pressure trip (-19 psig Based on discussions with Entergy [16], the RCIC operation for a in containment) 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> mission time requires RHR success in suppression pool (2) Required depressurization to meet PSP or HCTL cooling. For SBO, the operation of RCIC is limited by a number of requirements in the EOPs. factors:

These failure modes of RCIC are not discussed in - CST inventory Section 3.12 Item D nor in the T1 and SBO Success - Suppression pool temperature Criteria Tables in Appendix C. - Battery capacity

Attachment 1 to RBG-46922 Page 45 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs Importance to RI-ISI Gap #15 Because of the potential limitations, RCIC operation for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is (cont'd) credited. RCIC is not credited beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the SBO conditions.

The RCIC high turbine back pressure trip (even for cases of recirc seal LOCA) is expected to occur beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and therefore, is HPCS dependence on SPC for success under Large not limiting.

LOCA conditions needs to be identified in the Success NS: Not Significant.

Criteria Table. This is a documentation issue. The model is not being changed to address this item.

Gap #16 Limitations of codes are not discussed. SC-B4 hNS: Not Significant This is a documentation issue. The model is not being changed to address this item.

Gap #17 Comparative analysis with other plants is not performed for SC-B5 NS: Not Significant support. This is a documentation issue. The model is not being changed to address this item.

Gap #18 Documented. SC-Cl NS: Not Significant Basis is not clearly documented to allow a Peer Review Team to This is a documentation issue. The model is not being changed to independently assess the adequacy. address this item.

Gap #19 The RG 1.200 endorsement of the ASME PRA Standard has SC-C3 NS: Not Significant included a requirement to document the assumptions and sources The RBS PRA includes a substantial number of sensitivity of uncertainty associated with each PRA element. The NRC and calculations that demonstrate the range of uncertainties the industry are working together to clarify what this means and to -

• associated with specific assumptions and modeling uncertainties.

develop a structure that will satisfy these SRs. As of the These sensitivity studies are consistent with the expected performance of this self-assessment, this cooperative effort has NUREG-1855 approach to sensitivity analyses for evaluation of not been completed. modeling uncertainties.

Peer Review expectations and an edict from the BWROG to The RBS approach is to be determined once the new NRC/EPRI declare these SRs as Not Met will in general lead to "Not Met" guidance is available (e.g., NUREG-1855). However, the EPRI categorization by Peer Review Teams. RI-ISI process is defined such that model uncertainties will not unduly influence results.

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Gap #19 (cont'd) T . _

Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This

Attachment 1 to RBG-46922 Page 46 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap J Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #20 All alternate system alignments are not considered in the model SY-A5 = NS: Not Significant development and documented in the System Notebooks. See AS-B5a.

Gap #21 Cannot ascertain whether the system boundary is appropriately SY-A6 ° NS: Not Significant drawn for each system. This is a documentation issue. The model is not being changed'to address this item.

Gap #22 Component model boundaries are not explained in the component SY-A8 0 NS: Not Significant data notebook and are not shown to be consistent with the data This is a documentation issue. The model is not being changed to used. address this item.

Gap #23 System and functional success criteria are sequence dependent SY-A1 1 0 NS: Not Significant and time dependent but these are not explained in the PRA-RB- This is a documentation issue. The model is not being changed to 01-002S11. address this item.

Gap #24 Screening criteria from Supporting Requirement SY-A14 are not SY-A13

• NS: Not Significant referenced or discussed as they may be used to limit consideration This is a documentation issue. The model is not being changed to of low probability failure modes. address this item.

See PRA-RB-01-002S1 1.

Gap #25 No indication that this criteria is met regarding incorporation of SY-A14

• NS: Not Significant failure modes. This is a documentation issue. The model is not being changed to See PRA-RB-01 -002S11. address this item.

Attachment 1 to RBG-46922 Page 47 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap J Description of Gap to.Capability Category II Applicable SRs [ Importance to RI-ISI Gap #26 The following is developed as supporting documentation for River SY-A15 N/A Bend:

" A list of the PRA systems to consider for test and maintenance actions

" Procedures reviewed, the potential test and maintenance actions associated with the procedures, and the disposition of the action (screened or evaluated).

Not addressed are the following: NS: Not Significant 0 Identify T&M activities that require realignment of the system This task is found not to have significant impact on the PRA model outside its normal operational or stand by status. and results or the RI-ISI application. The pre-initiator HEP a Rules for identifying and screening test and maintenance analysis is supportive of Capability Category II applications. (Pre-actions from the PRA. initiator HEPs included as required.)

See PRA-RB-01-002S1 1.

Gap #27 The River Bend model does not include system dependencies on SY-A17 PA: Potentially Applicable accident progression including isolations and trips under severe accident conditions; e.g., RCIC back pressure trip; L8 trip on ref. Provide sensitivity by setting RCIC failure to 1.0 for-LOCAs, loss leg leakdown; MSIV closure interlock on low level and the bypass of DHR, and long term SBO sequences and MSIV closure for interface, turbine trip ATWS to 1.0.

See PRA-RB-01-002S1 1. Alternatively, update the model to address these two items.

Gap #28 Realistic functional requirements are not discussed to characterize SY-Al19 NS: Not Significant system operation. This is a documentation issue. The model is not being changed to address this item.

See PRA-RB-01-002S1 1. 1 6

Attachment 1 to RBG-46922 Page 48 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap I Description of Gap to Capability Category II I Applicable SRs t Importance to RI-ISI Gap #29 Repair appears to be used even if data is unavailable to support SY-A22

• PA: Potentially Applicable the repair probability. Perform a sensitivity evaluation to set all repair to 1.0 failure unless appropriate generic or plant specific data are available to Estimated "repair/recovery" values from NUREG/CR-4550 no support the quantification. (Excludes offsite AC and on-site AC longer meet the latest expectation for PRA established in the power recoveries.)

ASME PRA Standard. Therefore, these repair events are considered potentially significant in establishing the risk baseline The ones of specific interest are:

and not adequately supported by available data for: ORA-EDC4HRS

- EDC - ORA-PCS1HRS

- PCs - ORA-PCS4HRS

- DHR - ORA-DHRLT

- See PRA-RB-01-002S11 and Ref [16]. This was confirmed in Ref. [16].

Gap #30 See SY-B5. (Address room cooling with appropriate analysis. SY-B6 0 NS: Not Significant This is a documentation issue. The model is not being changed to address this item.

Gap #31 Inventories of air, water, and cooling are not treated explicitly in SY-B12 0 NS: Not Significant the model documentation. This is a documentation issue. The model is not being changed to Not discussed in a Dependency Notebook, individual System address this item.

Notebooks or Event Sequence Notebook.

Gap #32 System Notebooks do not provide this information. SY-C1

• NS: Not Significant The System Notebooks do not contain all the information This is a documentation issue. The model is not being changed to requested by the ASME PRA Standard. address this item.

Gap #33 The System Notebooks or Component Data Notebook do not SY-C2

• NS: Not Significant contain a significant fraction of the information requested by the This is a documentation issue. The model is not being changed to ASME PRA Standard. address this item.

Attachment 1 to RBG-46922 Page 49 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II [ Applicable SRs Importance to RI-ISI Gap #34 The system assumptions are clearly documented in PRA-RB SY-C3

• NS: Not Significant 002S11. The RBS PRA includes a substantial number of sensitivity No uncertainty evaluation is performed. calculations that demonstrate the range of uncertainties The RG 1.200 endorsement of the ASME PRA Standard has associated with specific assumptions and modeling uncertainties.

Thcluded RG 10enorsement tofdocu the t asM mpR tandad hus These sensitivity studies are consistent with the expected included a requirement to document the assumptions and sources NUREG-1855 approach to sensitivity analyses for evaluation of of uncertainty associated with each PRA element. The NRC and modeling uncertainties.

the industry are working together to clarify what this means and to develop a structure that will satisfy these SRs. As of the The RBS approach is to be determined once the new NRC/EPRI performance of this self-assessment, this cooperative effort has guidance is available (e.g., NUREG-1855). However, the EPRI not been completed. RI-ISI process is defined such that model uncertainties will not Peer Review expectations and an edict from the BWROG to unduly influence results.

declare these SRs as Not Met will in general lead to "Not Met" categorization by Peer Review Teams. EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #35 The following are not available as supporting documentation for HR-Al NS: Not Significant River Bend: This is a documentation issue. The model is not being changed to

• A list of the PRA systems to consider for test and maintenance address this item.

actions e Rules for identifying and screening test, inspection, and maintenance actions from the PRA 0 A list of procedures reviewed, the potential test and maintenance actions associated with the procedures, and the disposition of the action (screened or evaluated).

• T&M activities that require realignment of the system outside its normal operational or stand by status.

Gap #36 System Notebooks and System Manager Interviews offer the HR-A3 NS: Not Significant opportunity to include the identification of the work practices that In general, pre-initiators are not significant contributors to the could influence pre-initiators. BWR risk profile. In addition, the system manager interviews have marginal value in establishing the pre-initiators for the PRA. No additional action for RI-ISI is required.

Attachment 1 to RBG-46922 Page 50 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI Gap #37 The HEP screening process in the River Bend pre-initiator HR-B1

• NS: Not Significant evaluation is not identified consistent with ASME PRA Standard. This is a documentation issue. The model is not being changed to address this item.

Gap #38 Dependent pre-initiator HEPs are addressed where multiple trains HR-B2

• PA: Potentially Applicable or functions are affected. Perform a sensitivity calculation to set all miscalibrations of Miscalibration dependencies using Figure 2 of the PRA-RB multiple channels to 3E-5.

002S03 appears to be too optimistic regarding the assignment of multiple miscalibration errors because it does not reflect:

• Common measuring standards

• Common crews

• Common procedures .

The 1 E-8 used for miscalibration is judged to be non-conservative and not supported by the THERP or ASEP methods.

Miscalibration probabilities of 9E-16 in Table 2 are judged to be unsupportable and detract from the high quality of the RBS PRA.

Miscalibration of Low Rx Pressure Signals (LPCI/LPCS interlock) is listed as negligible. This is contrary to HR SR-B2 and is judged to be unsupportable and detract from the high quality of the RBS PRA.

Gap #39 Miscalibration is included. HR-C3 PA: Potentially Applicable Miscalibration dependencies using Figure 2 of the PRA-RB See HR-B2.

002S03 appears to be too optimistic regarding the assignment of multiple miscalibration errors because it does not reflect: (

• Common measuring standards

• Common crews

• Common procedures The 1 E-8 used for miscalibration is judged to be non-conservative and not supported by the THERP or ASEP methods.

Miscalibration probabilities of 9E-16 in Table 2 are judged to be unsupportable and detract from the high quality of the RBS PRA.

Miscalibration of Low Rx Pressure Signals (LPCI/LPCS interlock) is listed as negligible. This is contrary to HR SR-B2 and SR-C3 and is judged to be unsupportable and detract from the high quality of the RBS PRA.

Attachment 1 to RBG-46922 Page 51 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap [ Description of Gap to Capability Category II Applicable SRs Importance to RI-ISI Gap #40 Appropriate operator actions are included in HRA. HR-E2 PA: Potentially Applicable For actions other than "skill of the craft", the incorporation of CR Perform a sensitivity calculation to remove credit for all non-(Recovery HEPs) for non-proceduralized actions is generally not proceduralized recoveries and repairs unless there is explicit allowed without significant documentation support. documentation provided to justify the assigned "recovery".

(See SY-A22 for related SR.)

This is not met.

Gap #41 Plant specific MAAP calculations are not used to provide allowed HR-G4. No effect on the pipe rupture effects on CDF or LERF are identified.

times for crew response. As such, the RI-ISI can be performed effectively without the resolution of this SR to a higher Capability Category.

Thermal hydraulic analyses appropriate for River Bend are not used to set time available. SI: Specific Issue There is some possibility that Service Water Importance can be No generic analysis is presented or referenced to support influenced by this SR deficiency. Therefore, SW pipe segments allowable action times. need to be evaluated for this potential change in importance if this SR is resolved to be Capability Category I.

Interface of success criteria and plant specific calculations:

This can be resolved by using bounding CCDP and initiating HEP - BA-SSWINJ (Section 5.2.5 in PRA-RB-01-002S03) rupture frequencies if they could be determined.

allows 20 min. for SW alignment to prevent core damage.

See SC-A6 for sensitivity case to subsume this item.

This appears to be in need of a clear definition of core damage (based on a measurable parameter) as required by SR SC-B2 and a method to calculate the parameter (e.g.,

RBS MAAP calculation).

Neither of these two could be found.

The 20 min. time allowed for a DBA LOCA is judged to be significantly longer than any other BWR reviewed by the BWROG during the BWROG certification process.

Attachment 1 to RBG-46922 Page 52 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap J Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI Gap #42 Evidence that HEPs are reviewed for reasonableness is not HR-G6

• NS: Not Significant presented. This is a documentation issue. The model is not being changed to The calculated HEP derivations are not provided. address this item.

Section 5.2.6 (PRA-RB-01-002S03) ADS recovery. The HEP of 1.2E-5 is lower than considered possible using the cause based and THERP approaches.

RPT: For ATWS sequences subsumed into the ATWS model, manual RPT cannot be credited; in fact only RPT on high dome pressure can be credited.

SSW cross tie for ATWS and LOCA response would appear to be optimistic especially considering that no deterministic calculation is available to support its flow rate and timing as adequate as the sole RPV injection source.

Gap #43 The apparent separate treatment of HEPs, in-model recoveries, HR-I1 NS: Not Significant and ex-model recoveries needs to be better described and The treatment of the combination of HEPs within a given cutset is integrated. There needs to be an explanation of the combined assessed by Entergy. See Table 10 of the HRA Calculational HEP values that result from this treatment. Notebook. This evaluation was submitted by Entergy [16] and leads to the conclusion that the HEPs are adequately modeled for dependencies and may only require additional documentation to describe the process and display the results. This is a documentation issue. The model is not being changed to address this item.

Attachment 1 to RBG-46922 Page 53 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II I Applicable SRs Importance to RI-ISI Gap #44 The RG 1.200 endorsement of the ASME PRA Standard has HR-13 NS: Not Significant included a requirement to document the assumptions and sources The RBS PRA includes a substantial number of sensitivity of uncertainty associated with each PRA element. The NRC and calculations that demonstrate the range of uncertainties the industry are working together to clarify what this means and to associated with specific assumptions and modeling uncertainties.

develop a structure that will satisfy these SRs. As of the These sensitivity studies are consistent with the expected performance of this self-assessment, this cooperative effort has NUREG-1855 approach to sensitivity analyses for evaluation of not been completed. modeling uncertainties.

Peer Review expectations and an edict from the BWROG to The RBS approach is to be determined once the new NRC/EPRI declare these SRs as Not Met will in general lead to "Not Met" guidance is available (e.g., NUREG-1855). However, the EPRI dclategorezathee Ss arNtetwil inerl lRI-ISI process is defined such that model uncertainties will not categorization by Peer Review Teams. unduly influence results..

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology, used in the Code Case N-716 RI-ISI.

Gap #45 PRA-RB-01-002S05 Rev. 0 DA-Ala NS: Not Significant Component boundaries are not explicitly provided in the Data This is a documentation issue. The model is not being changed to Notebook. address this item.

Attachment 1 to RBG-46922 Page 54 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap J Description of Gap to Capability Category II ]- Applicable SRs [ Importance to RI-ISI Gap #46 Data Notebook provides no discussion or indication that coincident DA-C13

• NS: Not Significant maintenance was considered or evaluated. Since work practices call for a Protected Division philosophy, as To be consistent with SR DA-C13, the PRA should include an documented in ADM-0096 on-line maintenance procedure, cross-examination of coincident outage times for redundant equipment divisional maintenance unavailabilities would be limited to (both intra- and inter-system) and incorporate the results into the emergent situations and thus coincident maintenance modeling and documentation. However, it is judged that it is not unavailabilities would be expected to be negligible.

practical to model all potential combinations of coincident maintenance unavailabilities, and that a review of maintenance experience would not be sufficient to allow the prediction of the dominant risk contributor combinations.

As such, an approach to identify dominant risk contributor combinations based on knowledge of the accident sequences modeling, and model such combinations of coincident maintenance outages in the fault tree logic is judged prudent. A review of recent maintenance experience can be performed to identify events of coincident maintenance outages for these combinations to support probability estimation for the events.

Gap #47 Data Notebook provides no discussion or indication repair was DA-C14 9 PA: Potentially Applicable considered or included in PRA. See SY-A22.

Gap #48 Repair is not discussed in Data Notebook. Data is not sufficient to DA-C15 0 PA: Potentially Applicable support the ASME requirement. See SY-A22.

Gap #49 No indication from Data Notebook that past data is examined for DA-D7

• NS: Not Significant applicability of the data. Data used ranges from: This is a documentation issue. The model is not being changed to 1988 -2003 address this item.

1993 -2003 1998 -2003

-Provide confirmation that the data used is applicable given that plant modifications and procedures have significantly changed the as-built, as-operated plant.

Gap #50 No discussion or indication of repair actions in Data Notebook. DA-D8 See SY-A22.

(NewNRCSR)

Attachment 1 to RBG-46922 Page 55 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap 1 Description of Gap to Capability Category I1 Applicable SRs Importance to RI-ISI Gap #51 Assumptions are documented in Section 3.2 of PRA-RB DA-E3

• NS: Not Significant 002S05, Rev. 0. The RBS PRA includes a substantial number of sensitivity The RG 1.200 endorsement of the ASME PRA Standard has calculations that demonstrate the range of uncertainties included a requirement to document the assumptions and sources associated with specific assumptions and modeling uncertainties.

of uncertainty associated with each PRA element. The NRC and These sensitivity studies are consistent with the expected the industry are working together to clarify what this means and to NUREG-1855 approach to sensitivity analyses for evaluation of develop a structure that will satisfy these SRs. As of the modeling uncertainties.

performance of this self-assessment, this cooperative effort has not been completed. .The RBS approach is to be determined once the new NRC/EPRI Peer Review expectations and an edict from the BWROG to guidance is available (e.g., NUREG-1855). However, the EPRI PerReview texpecSsats ano aetdwlict gnrom thead to "Not MRI-ISI process is defined such that model uncertainties will not declare these SRs as Not Met will in general lead to "Not Met" unduly influence results.

categorization by Peer Review Teams.

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #52 Modules are used, but the requirements of QU-B9 are not QU-B9 0 NS: Not Sihnificant described in the RBS documentation. This is a documentation issue. The model is not being changed to PRA-RB-01-005 and PRA-RB-01-002S02. address this item.

Gap #53 No evidence of a sample of the significant accident QU-Dla 0 NS: Not Significant sequences/cutsets have been appropriately reviewed is presented. This is a documentation issue. The model is not being changed to PRA-RB-01-005 and PRA-RB-01-002S02. address this item.

Attachment 1 to RBG-46922 Page 56 of 68

- TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap ] Description of Gap to Capability Category II Applicable SRs J Importance to RI-ISI Gap #54 The results of the PRA model reflect the sequence models, system QU-Dlb Action models, success criteria, and the as-built, as-operated plant.

Provide the CDF and LERF along with its truncation.

Appendix F of PRA-RB-01-002S02 give the PRA Quant results at Ensure that the results are properly vetted.

1E-i1 /yr truncation to be 1.197E-5/yr. This appears to be significantly different than the NS: Not Significant reported point estimate of This is a documentation issue. The model is not being changed to 3.62E-6/yr. No explanation provided. address this item.

It also differs from Appendix D at 2.34E-6/yr.

Table 7 of PRA-RB-01-002S02 gives:

Truncation CDF(/yr) (yr) 3.62E-6 1E-11 3.20E-6 1E-10 (Cutset Truncation)

PRA-RB-01-005 and PRA-RB-01-002S02.

Gap #55 No evidence is found that dominant sequences are reviewed and QU-Dlc

• NS: Not Significant found consistent with model, plant, procedures, and mutually This is a documentation issue. The model is not being changed to exclusive file. address this item.

PRA-RB-01-005 and PRA-RB-01-002S02.

Gap #56 No evidence is found that non-dominant sequences are reviewed QU-D4

• NS: Not Significant and found appropriate. This is a documentation issue. The model is not being changed to PRA-RB-01-005 and PRA-RB-01-002S02. address this item.

Attachment 1 to RBG-46922 Page 57 of 68

- TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap__ Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI Gap #57 The RG 1.200 endorsement of the ASME PRA Standard has QU-E1 NS: Not Significant included a requirement to document the assumptions and sources The RBS PRA includes a substantial number of sensitivity of uncertainty associated with each PRA element. The NRC and calculations that demonstrate the range of uncertainties the industry are working together to clarify what this means and to associated with specific assumptions and modeling uncertainties.

develop a structure that will satisfy these SRs. As of the These sensitivity studies are consistent with the expected performance of this self-assessment, this cooperative effort has NUREG-1855 approach to sensitivity analyses for evaluation of not been completed. modeling uncertainties.

Reef Review expectations and an edict from the BWROG to The RBS approach is to be determined once the new NRC/EPRI declare these SRs as Not Met will in general lead to "Not Met" guidance is available (e.g., NUREG-1855). However, the EPRI dclategorezathee Ss ae Ntetwi ige r RI-ISI process is defined such that model uncertainties will not categorization by Peer Review Teams. unduly influence results.

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Attachment 1 to RBG-46922 Page 58 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs J Importance to RI-ISI Gap #58 See PRA-RB-01-005 and PRA-RB-01-002S02. QU-F4 s NS: Not Significant The RG 1.200 endorsement of the ASME PRA Standard has The RBS PRA includes a substantial number of sensitivity included a requirement to document the assumptions and sources calculations that demonstrate the range of uncertainties of uncertainty associated with each PRA element. The NRC and associated with specific assumptions and modeling uncertainties.

the industry are working together to clarify what this means and to These sensitivity studies are consistent with the expected develop a structure that will satisfy these SRs. As of the NUREG-1 855 approach to sensitivity analyses for evaluation of performance of this self-assessment, this cooperative effort has modeling uncertainties.

not been completed: The RBS approach is to be determined once the new NRC/EPRI guidance is available (e.g., NUREG-1855). However, the EPRI Peer Review expectations and an edict from the BWROG to RI-ISI process is defined such that model uncertainties will not declare these SRs as Not Met will in general lead to "Not Met" unduly influence results.

categorization by Peer Review Teams.

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #59 No evidence is found that the process for using CAFTA is part of QU-F5 e NS: Not Siqnificant CAFTA documentation. This is a documentation issue. The model is not being changed to address this item.

Gap #60 No evidence is found that the limitations of model are documented. QU-F6 e NS: Not Significant This is a documentation issue. The model is not being changed to address this item.

Gap #61 Plant damage states are not defined. LE-A5 0 NS: Not Significant This is a documentation issue. The model is not being changed to address this item.

The end states of the Level 1 do not appear to distinguish among possible plant damage states. They are all listed as CD (core damage) or OK. This may create difficulty in the Level 2 assessment because of the significant differences in RPV, containment, and plant conditions given the various accident sequence paths to core damage in Level 1.

PRA-RB-01-002S12 Rev. 0.

Gap #62 Scrubbing of fission products is not explicitly modeled with MAAP -. LE-C100 PA: Potentially Applicable

Attachment 1 to RBG-46922 Page 59 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap J Description of Gap to Capability Category II Applicable SRs Importance to RI-ISI or equivalent deterministic code. The use of NUREG/CR-6595 Perform a sensitivity calculation to eliminate all DF in the Auxiliary satisfies Capability Category I for pool scrubbing assessments. Building unless a calculation is available to support the assessment.

However, scrubbing by Aux. Bldg is credited by assumption in reducing the radionuclide releases. A Decontamination Factor (DF) is developed based strictly on engineering judgment.

Gap #63 ISLOCA frequency development is plant specific and considers LE-D3

• NS: Not Si-gnificant plant details. Discussion with Entergy [16] indicated that the low See IE-A2.

pressure pipe for LPCS and LPCI has been hydrostatically tested at normal operating pressures of the RPV (>1000 psig).

Therefore, there is high confidence that the pipe is capable of withstanding any potential high pressure condition that could result from a failure of the high pressure to low pressure interface valves.

Based on this plant unique resolution to the ISLOCA question, no additional sensitivity cases are needed.

However, the ISLOCA evaluation does not include critical lines that need to be addressed.

Attachment 1 to RBG-46922 Page 60 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs ] Importance to RI-ISI Gap #64 Containment isolation failure probability is judged not to be LE-D6

• NS: Not Significant adequate. No impact on CDF and minimal impact on LERF.

Containment bypass is modeled in the LERF model, as documented in PRA-RB-01-002S12 Section 3.1.3. This also Note that radionuclide scrubbing treatment in the Auxiliary documents that Containment Isolation is modeled as part of the Building is assessed in SR LE-C10.

RBS System Notebook. A more detailed containment isolation notebook (CIS-22) had previously existed in the 1990s that contained much more detail than is currently in PRA-RB 002S 11 / S20.

RBS LERF model does include gates/models for failure to isolate containment vent and purge (KJB-Z31, KJB-Z33), reactor floor drains (KJB-Z35), and reactor equipment drains (KJB-Z38).

Discussion in PRA-RB-01-002S12 implies a NUREG/CR-6595 LERF model, such as RBS, does model whether or not the Suppression Pool provides scrubbing of releases. (See LE-C10.)

In addition, Entergy stated in Ref. [16] that the LERF model for RBS uses the assumptions from NUREG/CR-6595.

The containment isolation notebook needs to be reissued and updated (CIS-22).

The containment isolation analysis appears to be missing treatment of:

• Pre-existing DW and containment failures PRA-RB-01-002S12 Rev. 0.

Gap #65 Dominant contributors to LERF are not provided. LE-Fla 0 NS: Not Significant Attachment 6, PRA-RB-01-002S12 Rev. 0. This is a documentation issue. The model does not need to be changed to address this item.

Gap #66 Dominant contributors to LERF are not discussed for LE-Fi b 0 NS: Not Significant reasonableness. This is a documentation issue. The model does not need to be Attachment 6, PRA-RB-01-002S12 Rev. 0. changed to address this item.

Attachment 1 to RBG-46922 Page 61 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap I Description of Gap to Capability Category II -Applicable SRs Importance to RI-ISI Gap #67 Compliance with SR LE-F2 includes the LERF analysis and LE-F2 NS: Not Significant associated documentation which incorporates: The RBS PRA includes a substantial number of sensitivity

• Quantitative sensitivity studies of the LERF analysis to reflect calculations that demonstrate the range of uncertainties variations in the significant LERF contributors associated with specific assumptions and modeling uncertainties.

The RG 1.200 endorsement of the ASME PRA Standard has These sensitivity studies are consistent with the expected Thcluded RG 10enorsement tofdocumt the asM mpR tandad hus NUREG-1855 approach to sensitivity analyses for evaluation of included a requirement to document the assumptions and sources modeling uncertainties.

NRC and of uncertainty associated with each PRA element. The the industry are working together to clarify what this means and to The RBS approach is to be determined once the new NRC/EPRI develop a structure that will satisfy these SRs.\As of the guidance is available (e.g., NUREG-1855). However, the EPRI performance of this self-assessment, this cooperative effort has RI-ISI process is defined such that model uncertainties will not not been completed. unduly influence results.

Peer Review expectations and an edict from the BWROG to EPRI [15] has reviewed the ASME PRA Standard Supporting declare these SRs as Not Met will in general lead to "Not Met" Requirements for their applicability in ensuring the technical categorization by Peer Review Teams. quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Attachment 1 to RBG-46922 Page 62 of 68 TABLE 1 STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap Description of Gap to Capability Category II Applicable SRs; Importance to RI-ISI Gap #68 The RG 1.200 endorsement of the ASME PRA Standard has LE-F3. NS: Not Significant included a requirement to document the assumptions and sources The RBS PRA includes a substantial number of sensitivity of uncertainty associated with each PRA element. The NRC and calculations that demonstrate the range of uncertainties the industry are working together to clarify what this means and to associated with specific assumptions and modeling uncertainties.

develop a structure that will satisfy these SRs. As of the These sensitivity studies are consistent with the expected performance of this self-assessment, this cooperative effort has NUREG-1855 approach to sensitivity analyses for evaluation of not been completed. modeling uncertainties.

Peer Review expectations and an edict from the BWROG to The RBS approach is to be determined once the new NRC/EPRI declare these SRs as Not Met will in general lead to "Not Met" guidance is available (e.g., NUREG-1855). However, the EPRI categorization by Peer Review Teams. RI-ISI process is defined such that model uncertainties will not unduly influence results.

EPRI [15] has reviewed the ASME PRA Standard Supporting Requirements for their applicability in ensuring the technical quality of RI-ISI risk-informed decisions. Based on the EPRI evaluation [15], this Supporting. Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #69 Accident sequences, binning, and plant damage status are LE-G1 0 NS: Not Significant marginally documented. Additional documentation is judged to be This is a documentation issue. The model is not being changed to necessary. This includes the nodal fault trees and their address this item.

assumptions.

Gap #70 Additional documentation is judged to be necessary. This includes LE-G3

• NS: Not Significant the nodal fault trees and their assumptions. This is a documentation issue. The model is not being changed to I_ address this item.

Attachment 1 to RBG-46922 Page 63 of 68 TABLE I STATUS OF "NOT MET" SUPPORTING REQUIREMENTS OF THE ASME PRA STANDARD TO SUPPORT CODE CASE N-716 FOR RI-ISI Gap ] Description of Gap to Capability Category II Applicable SRs [ Importance to RI-ISI Gap #71 The Level 2 Notebook and the PSA Summary Notebook provide LE-G4

• NS: Not Significant the following: The RBS PRA includes a substantial number of sensitivity

" Quantitative sensitivity studies of the LERF analysis to reflect calculations that demonstrate the range of uncertainties variations in the significant LERF contributors associated with specific assumptions and modeling uncertainties.

These sensitivity studies are consistent with the expected

• Assumptions that could impact LERF results NUREG-1855 approach to sensitivity analyses for evaluation of The RG 1.200 endorsement of the ASME PRA Standard has modeling uncertainties.

included a requirement to document the assumptions and sources of uncertainty associated with each PRA element. The NRC and The sensitivity cases are judged to be determined once the new the industry are working together to clarify what this means and to NRC/EPRI guidance is available (e.g., NUREG-1 855). However, develop a structure that will satisfy these SRs. As of the the EPRI RI-ISI process is defined such that model uncertainties performance of this self-assessment, this cooperative effort has will not unduly influence results.

not been completed.

EPRI [15] has reviewed the ASME PRA Standard Supporting Peer Review expectations and an edict from the BWROG to Requirements for their applicability in ensuring the technical declare these SRs as Not Met will in general lead'to "Not Met" quality of RI-ISI risk-informed decisions. Based on the EPRI categorization by Peer Review Teams. evaluation [15], this Supporting Requirement does not need to be met in order to adequately support the EPRI RI-ISI application methodology. This evaluation is also judged to apply to the methodology used in the Code Case N-716 RI-ISI.

Gap #72 The quantitative definition used for significant accident progression LE-G6 NS: Not Significant sequence is not included. This is a documentation issue. The model is not being changed to address this item.

Attachment 1 to RBG-46922 Page 64 of 68 TABLE 2 DISCUSSION OF POTENTIALLY APPLICABLE GAPS TO REGULATORY GUIDE 1.200 Gap Description of Gap to Capability Category II Applicable Disposition SRsa Initiators appropriately categorized and plant specific features accounted for.

Internal flooding initiators to be completed by Entergy.

The implementation of the ISLOCA evaluations in PRA-RB 002S08 has assumed that any 4 valves in a line qualify to allow screening a line from consideration. This is judged to be inconsistent with NSAC-154 and typical PRA practice.

The isolation valves that are to be counted must:

(1) be able to close against the differential pressure or (2) must be closed as their normal position The lines screened are the 3 LPCI and 1 LPCS injection lines.

This gap is inherently addressed through the Internal Flooding PRA, which Low pressure rated pipe in the LPCI, SDC, and LPCS systems will determine the risk significance of the individual ECCS/etc.. line has been hydrostatically tested at rated RPV pressure.

segments. Use of the "Sensitivity" cases which account for the Discussion with Entergy [16] indicated that the low pressure pipe Gap #1 IE-A2 contribution of component rupture frequencies in addition to the EPRI pipe for LPCI and LPCI has been hydrostatically rupture frequencies will ensure that this ECCS / RCIC / FWS / RWCU tested at normal operating pressures of the RPV (>1000 psig).

piping will be appropriately characterized as part of the RBS Code Case N-Therefore, there is high confidence that the pipe is capable of 716 RIISI program.

withstanding anylISLOCA condition. Based on this plant unique resolution to the ISLOCA question, no additional sensitivity cases are needed for the LPCI, SDC, or LPCS lines.

Breaks outside containment in high energy lines beyond the 2nd isolation valve (Main Steam, FW, HPCS, RWCU, RCIC) are also in need of evaluation to ensure these are properly accounted for.

(See IE-C4)

Reference Leg leak-down is an initiating event that can compromise multiple systems. This initiator should be identified for disposition. A loss of Ref. Leg is not assessed.

(See SLI-8211 [1], SL18218 [2], SL18221 [3] for typical approaches used in BWR PRAs).

I h

Attachment 1 to RBG-46922 Page 65 of 68 TABLE 2 DISCUSSION OF POTENTIALLY APPLICABLE GAPS TO REGULATORY GUIDE 1.200 Gap Description of Gap to Capability Category II Applicable Disposition SRs River Bend includes failure of RCIC on loss of NPSH and other Level 8 trip. River Bend has incorporated a reference leg backfill system which would mitigate against a reference leg leak. The MSIV closure interlock is of lower importance to the River Bend PRA model since RBS has motor-The River Bend model does not include system dependencies on driven, vice steam-driven, Main Feedwater Pumps. The RBS PRA fault accident progression including isolations and trips under severe tree models long-term RCIC failure on a loss of containment heat removal accident conditions; e.g., RCIC back pressure trip; L8 trip on ref. to account for the RCIC turbine trip on high exhaust back pressure. We Gap #27 leg leakdown; MSIV closure interlock on low level and the bypass SY-A17 believe that many of the issues have been addressed in the model and the interface. gap may be primarily documentation in nature. Sensitivity cases on RCIC will be considered during the Sensitivity and Uncertainty Calculation base See PRA-RB-01-002S1 1. model update. Thus, upon review, this gap is considered to not impact the use of the River Bend PRA for support of RI-ISI applications and is considered to be a documentation or very small significance issue that will be reviewed and addressed during the next PRA model update.

Repair appears to be used even if data is unavailable to support the repair probability.

Estimated 'repair/recovery" values from NUREGICR-4550 no A sensitivity was performed with these recoveries set to 1.0. This run longer meet the latest expectation for PRA established in the showed that CDF only increased by 26% (primarily due to the.Power ASME PRA Standard. Therefore, these repair events are Conversion System (PCS) recovery), but this is for the baseline internal events CDF. Internal Flooding Analyses are insensitive to offsite power Gap #29 considered potentially significant in establishing the risk baseline SY-A22 recovery actions since internal flooding scenarios do not lead to Loss of and not adequately supported by available data for:

a EDC Offsite Power related initiators. We believe that the impact of this issue on EDCS the internal flooding results is not significant enough to change any

• PCs conclusions made in this submittal

• DHR See PRA-RB-01-002S11.

Dependent pre-initiator HEPs are addressed where multiple trains Appendix B of the River Bend HRA calculation addresses the use of these or functions are affected. low miscalibrations for Reactor Level and Pressure. The HEPs for Reactor Level and Reactor Pressure miscalibrations are considered negligible Miscalibration dependencies using Figure 2 of the PRA-RB because there are multiple sets of these pressure and level signals 002S03 appears to be too optimistic regarding the assignment of (narrow range and wide range) for multiple system actuations. If all of the Gap #38 multiple miscalibration errors because it does not reflect: HR-B2 Reactor Level sensors for RCIC actuation are miscalibrated, the operators

• Common measuring standards would notice the miscalibration based on a comparison to level sensors for

• Common crews HPCS, RPS, etc. None of these transmitters are calibrated at the same

• Common procedures time and use different procedures. In addition, the level and pressure The 1 E-8 used for miscalibration is judged to be non-conservative sensors would have to be grossly miscalibrated to fail to actuate ECCS and not supported by the THERP or ASEP methods. components in a manner that would impact the success criteria.

Attachment 1 to RBG-46922 Page 66 of 68 TABLE 2 DISCUSSION OF POTENTIALLY APPLICABLE GAPS TO REGULATORY GUIDE 1.200 Gap Description of Gap to Capability Category II I Applicable SRs Disposition Therefore, River Bend has justified the values used for miscalibration of Miscalibration probabilities of 9E-16 in Table 2 are judged to be these instruments.

unsupportable and detract from the high quality of the RBS PRA.

Further, Sensitivity analyses were conducted using the recommended Miscalibration of Low Rx Pressure Signals (LPCI/LPCS interlock) 3.OE-05 value for miscalibration. These analyses demonstrated that there is listed as negligible. This is contrary to HR SR-B2 and is judged was no change in the ocntribution of flooding events to CDF or LERF.

to be unsupportable and d`tract from the high quality of the RBS PRA.

Appropriate operator actions are included in HRA.

For actions other than "skill of the craft", the incorporation of CR The River Bend PRA has only one Control Room human action event for manually opening SWP-AOV599 for non-SBO events. The operators are Gap #40 (Recovery HEPs) for non-proceduralized actions is generally not HR-E2 trained on this valve for SBO events. This action is a "skill of the craft" allowed without significant documentation support, based on the simplicity of the task and the training for the SBO sequence.

This is not met.

To determine the potential impact of containment bypass, a sensitivity case was run eliminating all credit for Auxiliary Building scrubbing in the LERF model. Event L2-ABSCRUB is set to "TRUE" vice assuming a 0.25 value. This resulted in a 205% increase in the LERF calculated due to Scrubbing of fission products is not explicitly modeled with MAAP flooding scenarios, to 6.31 E-08/year. Even with this conservative or equivalent deterministic code. modeling, the maximum LERF contributor (Feedwater breaks in the Main Steam Tunnel) had a calculated LERF of 3.62E-08, continuing to meet the 10-7 N-716 criteria. Thus, LERF results were demonstrated to meet the 10-Containment bypass accident sequences are not explicitly 7 criteria even with conservatively crediting no scrubbing due to the Gap #62 modeled. LE-C10 Auxiliary Building. In the determination of CLERP values greater than 10-5 for ASME Class II piping, no other scenarios with Class 2 pipe had CLERP Scrubbing by Aux. Bldg is credited by assumption in reducing the values greater than 3E-06, thus revisions to the auxiliary building radionuclide releases. A Decontamination Factor (DF) is scrubbing model would be expected to have no impact on the identification developed based strictly on engineering judgment. of high CLERP Class 2 lines. Also, the 1E-08/year EPRI per system LERF criteria and the total ARLERF would not be challenged by this conservative modeling (see Section 3.4.1). There is no impact on CDF associated with this modeling issue. Thus, this issue is concluded to have no impact on the use of the RBS PRA for RI-ISI applications.

to RBG-46922 Page 67 of 68 TABLE 3 DISCUSSION OF OPEN "A" AND "B" PEER REVIEW FACTS AND OBSERVATIONS (F&O'S)

PEER REVIEW ELEMENT ID / POSSIBLE RESOLUTION IMPACT LEVEL OF SIGNIFICANCE I OBSERVATION S-29 B AOP-0050 (Attachment 7) Make PRA model consistent An HRA was performed that evaluated the probability that the Issues related to SBO and Suppression Pool with AOP-0050. Consider crew would not perform properly Attachment 7 of the.-

Temperature Determination are the following: whether feedback to AOP Abnormal Procedure for Station Blackout (AOP-050). The

1. Asserts Suppression Pool Temperature from PSA is desirable to HEE was included in the fault tree and a sensitivity analysis (SP/T) Indication is unavailable in Control Room for make guidance on SBO was run at 1 E-1 3 truncation. The Birnbaums were compared an SBO. response optimized to avoid to the model that does not include the HEE. None of the

2. States Design Calculations Show SP/T HCTL and maintain RCIC Birnbaums increased by a factor of 3. Typically they does not exceed HCTL for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, operability, increased by less than 1%.

3. , AOP-0050 directs using local temperature Therefore it was concluded that the consequences ofi":,

measurements. improper operator action during local temperature Related operator actions are not included in the measurements directed by AOP-0050, Attachment 7,; have an HRA for control of RPV pressure below HCTL for insignificant effect on MSPI calculation basis (or RIISI)w SBO sequences using RCIC.

The assertion that HCTL is not exceeded for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in an SBO is contrary to the design calculation G13.18.12.4

• 4 for the no operator action to depressurize case, which would be the situation if no other guidance is available. It is believed normal training is in place to allow initial RPV pressure reduction to 500 psig and this would increase margin to HCTL. This action is not included in PSA model.

IE-6 B Special initiating eventswere discussed in the Include consideration of Level Break outside containment is not included in the current initiating event notebooks. However the following 2 and consider the lower model. However, BOC, MSL breaks, and FWL breaks are initiating events are believed to be incorrectly truncation that may be used expected to be a minimal risk impact to the CDF. The screened from the quantification process: Breaks in the application of PSA in frequency of an unisolated large or medium LOCA would be outside containment(BOC) deciding on the retention of about 1.OE-8/yr with no mitigation efforts:

1)Main steam line, 2)Feedwater lines, 3)RCIC & special initiators. This could This is calculated with (INI-A+INI-S1)*MSIV CCF to close.

RWCU Lines be performed qualitatively if The Large LOCA frequency is 3.2E-5/yr. The Medium LOCA The analysis is completely adequate and information is available to frequency is 3.32E-5/yr. The common cause failure of MSIVs appropriate for an IPE study or a study comparable support diverse and to close is 1.54E-4. Therefore, the BOC frequency is 1.OE-to RBG-46922 Page 68 of 68 to NUREG-1150. Because of the potential for redundant equipment 8/yr. This is calculated without credit for recovery actions.

Level 2 impacts (e.g., LERF), there is not a good available to mitigate the Smaller breaks and leaks are significantly less important reason presented to eliminate these from initiator. because smaller breaks and leaks damage less equipment.

quantification for a PSA that may be used for An unisolated break outside containment does not lead detailed applications. These sequences emphasize directly to core damage. The impact of this event is tied to an the need for the isolation function and the need to operator action for keeping the water level below the MSIV address the consequences of isolation failures. lines to prevent inventory loss other than from steam. HPCS Applications involving the isolation function and and low pressure injection sources would be available to potential large offsite releases could be affected. maintain level. This could become an inventory issue since NUREG-1602 (DRAFT) has indicated similar steam would go to the turbine building instead of back to the concerns by stating that a steam generator event suppression pool.

may have a relatively low contribution to the total When failure of the injection systems and failure of MSIVs is CDF but may constitute a significant fraction of total considered, the conditional core damage probability is 1 E-3 or large early releases. less. When these failures are in sequences with a BOC initiator, the CDF is 1E-1 1/yr or less. The impact of failure to model the BOC initiator is negligible and would have very limited impact on the importance of systems/components part of MSPI.

This impact is not expected to exceed the LOCA CDF contribution. The ISLOCA evaluation addresses RCIC and RWCU line breaks.

For Internal Flooding, these breaks outside containment have been explicitly modeled, thus this F&O has no impact.upon use of the RBS PRA for the RIISI application.

Attachment 2 to RBG-46922 Licensee-Identified Commitments to RBS-46922 Page 1 of 1 LICENSEE-IDENTIFIED COMMITMENTS The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE COMMITMENT (Check one) SCHEDULED COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE Request for Alternative CEP-ISI-007 X Upon NRC pertaining to the application of ASME Code approval of this Case N-663 will be withdrawn for use at RBS request for upon NRC approval of the RISB Program alternative submittal.

Consistent with previously approved RI-ISI X Within one (1) submittals, RBS will calculate coverage and year after the use additional examinations or techniques in end of the the same manner it has for traditional interval Section Xl 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 submitted per the guidance of 10 CFR 50.55a(g)(5)(iv).

Upon approval of the RISB Program, X Upon NRC procedures that comply with the guidelines approval of this described in EPRI TR-1 12657 will be request for prepared to implement and monitor the alternative program.